Technical Meeting on Long-Pulse Operation of Fusion Devices
Board Room A
IAEA Headquarters
KEY DEADLINES
30 June 2022 Deadline for submission of abstracts through IAEA-INDICO for regular contributions
30 June 2022 Deadline for submission of Participation Form (Form A), and Grant Application Form (Form C) (if applicable) through the official channels
29 July 2022 Notification of acceptance of abstracts and of assigned awards
Controlling fusion plasma for long periods, while gaining experience in steady-state and/or long-pulse operation with active cooling systems that can maintain the plasma facing components at a stable temperature, is essential for the success of ITER and fusion demonstration power plants.
To facilitate the coordination on these challenges, the IAEA and the International Energy Agency (IEA) have established – in 2020 – a network for Coordination on International Challenges on Long duration OPeration (CICLOP). The objectives of the CICLOP group are to promote activities, collect and disseminate information on the physics and engineering issues of long-pulse operation for tokamak and stellarator facilities, by sharing best practice, operational procedures, experimental data, simulation programme and coordinating experiments between the fusion-related IEA Technology Collaboration Programmes in close cooperation with the IAEA activities in the same field, through a series of Technical Meetings on Long-Pulse Operation of Fusion Devices.
Objectives
The event aims to review, discuss and address scientific and engineering issues related to steady-state and long-pulse operation of fusion devices, which are essential for ITER and future fusion reactors.
Target Audience
The event aims to bring together junior and senior fusion scientists, plasma physicists (theoreticians, modellers and experimentalists) and engineers in order to cover the physics and engineering issues of long-pulse operation for tokamak and stellarator facilities.
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Welcome and IntroductionConveners: Matteo Barbarino (International Atomic Energy Agency), Xavier LITAUDON (CEA)
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LPO sessionConvener: Christopher Holcomb (Lawrence Livermore National Laboratory)
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Exploration of long-pulse and steady-state operations in ITER
The ITER long-pulse and steady-state operations, foreseen in the ITER Research Plan [IRP, 2018 ITER technical report, ITR-18-03], are important steps towards exploration of reactor relevant tokamak operation and research. A few key research areas, such as the operational space with heating mixes and external current drive, access to target plasma states with profile tailoring and control, and plasma MHD stability, will be studied first in the Pre-Fusion Power Operation (PFPO) phase, minimizing impacts on the lifetime of the CS coils. The operational space, scenario recipes, and control techniques will be re-established with high performance fusion plasmas in the Fusion Power Operation (FPO) phase, which includes long-pulse operational aspects, such as the heat loads on the plasma facing components and continuous operation of various tokamak components. In this work, the operational space of PFPO long-pulse plasmas has been explored including scans of plasma density and heating mixes, and then extended for FPO long-pulse and steady-state operations including options for a heating and current drive (H&CD) upgrade. The foreseen options for an H&CD upgrade do not include a Lower Hybrid current drive system, and are therefore without strong far off-axis current drive. The optimization of the ideal MHD stability has been performed using the potential upgrade H&CD options. Operational recipes for accessing target plasma states and achieving a long pulse duration have been explored using integrated scenario modelling. The heat loads onto the plasma facing components have also been assessed using the SOLPS-ITER code, including potential mitigation using Neon impurity seeding. The analyses carried out in this work demonstrate the feasibility of achieving long-pulse H/He operation in PFPO and steady-state DT operation in FPO, in accordance with the ITER Research Plan objectives.
Speaker: Sun Hee KIM (ITER Organization) -
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Progress on Long-pulse Steady-state High Performance Plasmas on EAST
Significant progress has been achieved on EAST in the development of long-pulse steady-state advanced plasmas, and in the understanding of the related scientific and technical issues in support of ITER and future fusion reactors.
A thousand-second time scale (~1056s) fully non-inductive plasma has been achieved on EAST at the end of 2021 with poloidal beta ~1.5, a normalized confinement factor H89 ~1.3 at DN configuration with full metal wall using an actively cooled ITER-like tungsten divertor. The total injected energy into the plasma is ~1.73 GJ with Radio Frequency (RF) power. Key technical and scientific challenges have been addressed for steady-state operation. A robust plasma control is demonstrated to keep the equilibrium with good accuracy overcoming the challenge of drift in magnetic measurements over long pulses. An improved loop voltage control is key to sustain fully non-inductive CD. Meanwhile, new lower divertor can significantly mitigate the power exhaust challenge, enabling the handling of large divertor heat fluxes, up to 10 MWm-2, preventing impurities (particularly tungsten from the divertor) from contaminating and cooling down the plasma core, and maintaining good particle exhaust to ensure that the plasma density does not rise in an uncontrolled way. Taking advantage of synergistic effects has enabled fully non-inductive operation with RF driven current fraction fRFCD~70% and bootstrap current fraction fBS~30%: electron heating using on-axis Electron Cyclotron Heating (ECH) enhances the heating and current drive from Lower Hybrid Wave (LHW). A self-regulated system resulting from multiscale interaction between core MHD instabilities and electron temperature gradient (ETG) induced turbulence also contributes to the sustainment of the long pulse steady-state regime.
In support of steady-state high-performance operation for future fusion reactors, a long-pulse fully non-inductive regime with higher bootstrap current fraction and higher fusion performance is further explored in 2022. Recently, with the improved flexibility and capabilities, a duration of 310s H-mode plasma (H98y2>1.3, ne/nGW>0.6, fBS>50%) has been demonstrated at high density with zero torque on EAST. Higher density and poloidal beta increase the bootstrap current fraction and self-consistently broaden the current density profile, leading to a further increase in confinement. The key physics processes (e.g. RF synergy, confinement, transport, particle and heat exhaust etc.) of this steady-state regime will be illustrated. Extension of fusion performance with enhanced capabilities on EAST can offer unique contributions towards the successful operation of ITER.Speaker: Juan Huang (CnIPPCAS)
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10:35
Coffee
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LPO sessionConvener: Christopher Holcomb (Lawrence Livermore National Laboratory)
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Challenges on Long Pulse Operation in LHD
The Large Helical Device (LHD) started its operation in 1998. One of the main objectives of the LHD project is the comprehensive study for the steady-state operation towards fusion reactors. Since the plasma current is not essential for LHD/stellarator devices to confine plasma, it is free from intensive efforts to drive and sustain plasma current. However, any other issues necessary for achieving long pulse operation are still left to LHD/stellarators. In this conference, summary of steady-state studies for more than 20 years in LHD are presented, focusing on common issues both in LHD/stellarators and tokamaks.
In LHD, ultra-long pulse discharges more than 47 minutes were obtained. The plasma was heated and sustained by ECH and ICH whose power was about 1.2 MW, and total energy of 3.36 GJ was injected into the plasma, which is the world record between stellarators and tokamaks. Active plasma control via plasma heating and fuelling/pumping was the key to the steady-state operation. Fast interlock system and the real-time feedback control system of ICH/ECH and the gas puffing worked successfully in accordance with changes of the plasma wall interaction (PWI) effects. A stable plasma of electron and ion temperatures of about 2 keV and electron density of 1.2E19m-3 were maintained. In the operation, when the plasma density abruptly increased, RF power was quickly boosted up, then plasma temperature soon recovered. For optimization of the heating efficiency of ICH, minority ion ratio was also controlled, which resulted in the spark mitigation and consequently in the avoidance of the impurity accumulation.
The global particle balance during long duration helium plasma was analyzed, which revealed key physics of particle behavior between plasma and the vacuum vessel wall. It was observed, in LHD with graphite divertor plates and with stainless-steel vacuum vessel wall, that there exists dynamic change of the helium wall retention. By quantitative analyses of particle balance, it was found that the change could be attributed to two kinds of helium reservoir and those retention capabilities depending on temperature, i.e., graphite divertor plates and co-deposition layers on the stainless-steel vacuum vessel wall. In detailed analyses of the co-deposition layer with the TEM observation, it was found that the layer consists of mixed materials of carbon and iron. Another wall retention effect by the boron powder injection was recently observed, which contributes not only to edge particle control but also to core plasma performance.
Long duration discharges were terminated mainly by excess outgassing from highly heated plasma facing components, e.g., divertor plates and/or armors. In order to lower the surface temperature of the divertor plates, the actively cooled divertor component consisting of tungsten target plate and copper heat sink has been developed, which can withstand steady-state heat flux of about 20 MW/m2. Another engineering development about cryogenic pump which can be used in the neutron irradiated environment like ITER and/or DEMO will also be introduced at the conference.Speaker: Tomohiro Morisaki (National Institute for Fusion Science) -
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The stellarator W7-X on the way to long pulse operation
The Wendelstein 7-X stellarator has a super-conduction coil system, to prove the steady state capabilities of optimized stellarators. After different steps with a limiter (OP1, starting December 2015) and two runs with inertially cooled divertors (OP1.2a, OP1.1b, up to December 2018) the device was completed to fulfill the steady-state capabilities:
- A High-Heat Flux Divertor (HHF) for steady state loads of 10 MW. Its surfaces are covered with CFC (carbon).
- To derive this power in steady-state, the Divertor and many other in-vessel have to be water-cooled continuously.
- To enhance the particle pumping from the divertor, 10 cryo-pumps have been installed behind the divertor targets.
- The divertor targets have to be observed continually components, in order not to be overloaded. Therefore IR-observation of all divertor targets is required, to end the heating of a discharge in time. This is the main safety measure to avoid problems with the HHF-Divertor.
- Further diagnostic system have been build up the coming phases.
The completing phase of W7-X took longer than expected, also due to the COVID-pandemie, but was finished in December 2021. Since then, the commission is running, especially the filling and balancing of the more than 600 cooling circuits. First Plasma experiments are scheduled to take place in November 2022.
In addition to these hardware update, also the organization of the plasma operation, the data acquisition and the control system were further improved.
This report will look backward to the OP 1.2a/b with plasma operation in an inertial divertor, and the corresponding restrictions, limiting the energy (Power times plasma duration) deposited in a discharge. Nevertheless, the effect of the optimized configuration of W-7X has already been proven.
After discussing the completing phase with the hardware discussed above we will present the first results from the commissioning and operation phase.Speaker: Hans-Stephan Bosch (Max-Planck-Institute for Plasma Physics) -
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Long pulse operation in a tungsten environment: achievements and work plan for WEST
The WEST tokamak has recently fully completed the installation of its ITER-like actively cooled divertor, and it is now ready to pursue integrated investigations for the qualification of the ITER divertor components (named PFU for Plasma Facing Units) as well as for operating high performance plasmas in a full tungsten environment. A comprehensive diagnostic survey tracks the evolution of the PFUs, and a post-mortem work plan characterizes the material transformation after exposure [Martin2021,Richou2021], including on-purpose local melting, erosion/redeposition and emissivity evolution [Gaspar2020]. Real-time protection via Infrared cameras is a topic of intense research due to the complexity of disentangling direct and reflected signals [Talatizi2021], and converting them into real temperatures with the large spatial and temporal variation of the tungsten emissivity. The scenario development activity in this environment is twofold. First, it aims at obtaining the intense heat fluxes, both stationary and transient, that are required to study the ageing of the ITER PFUs. In this aspect, noticeable achievements are the realization of discharges with a divertor heat flux reaching 5 MW/m2 for less than 5 MW of injected power, up to 100 MW/m2 on specifically designed tungsten blocs for melting experiments [Corre2021], and the discovery of an Optical Hot Spots issue due to PFU inter-spacing [Diez2020,Gunn2021]. The second aspect of scenario development aims at investigating integrated high performance plasma scenarios compatible with the stress on the tungsten walls (ageing issue) and the tungsten contamination (transport issue). In this respect, long pulse discharges (up to 55s so far) demonstrated a good particle and heat exhaust capability [Loarer2020], and transitions to H-mode confinement regime were maintained for up to 4s, close to the threshold power [Vermare2021]. The limitation foreseen for longer pulses relies on MHD limits, and for H-mode scenarios, it will be necessary to gain margins by reducing the radiated power enhanced after the pedestal formation, or increasing the injected power. For these two aspects and several others, the installation of a 3 MW ECRH system at the end of 2023 will open the operational space and help addressing essential issues for the development of integrated scenario for a fusion reactor using tungsten plasma facing components.
[Martin2021] C Martin et al 2021 Phys. Scr. 96 124035
[Richou2021] M Richou et al 2021 Phys. Scr. 96 124029
[Gaspar2020] J Gaspar et al 2022 Nuclear Fusion https://doi.org/10.1088/1741-4326/ac6f68
[Talatizi2021] C Talatizi et al. 2021 Fusion Engineering and Design 171 112570
[Corre2021] Y Corre et al 2021 Phys. Scr. 96 124057
[Diez2020] M Diez et al 2020 Nuclear Fusion 60 054001
[Gunn2021] JP Gunn et al 2021 Nuclear Materials and Energy 27 100920
[Loarer2020] T Loarer et al 2020 Nucl. Fusion 60 126046
[Vermare2021] L. Vermare et al 2022 Nucl. Fusion 62 026002Speaker: PATRICK MAGET (CEA)
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12:45
Lunch
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LPO sessionConvener: Stefano Coda (CRPP-EPFL)
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Overview of Recent and Planned DIII-D Research to Develop Steady-State Tokamak Operation for Fusion Energy
C. Holcomb for the DIII-D Team
Lawrence Livermore National LaboratoryDIII-D is focused on providing the scientific basis of high fusion performance, noninductively-sustained tokamak operation in ITER and pilot plants that will set the stage for commercial energy production. This presentation will highlight recent DIII-D research progress investigating core plasma scenarios ranging from “high beta hybrid” to “high qmin”. These studies have targeted improved understanding of limits to high normalized-beta steady-state operation, including ideal and resistive MHD instabilities, heat and particle transport, and current drive. For example, application of more off-axis NBI power to elevated-qmin discharges has in some cases produced broader profiles with higher ideal-MHD betaN limits and more classical fast ion transport. Studies have begun to push core scenarios to more reactor-relevant conditions, such as lower rotation, higher Te/Ti, higher density, lower collisionality, etc, and to assess compatibility and integration with boundary requirements such as no Type I ELMs and highly radiative detached divertors. High-betaN hybrid scenario plasmas are a case in point, where studies have shown that at high heating power lowering torque reduces normalized confinement but raising density can increase it. Plans to expand DIII-D’s capabilities for reactor-relevant studies will be discussed. These include significant increases in heating and current drive power and flexibility in the form of new ECH lines, a high harmonic fast wave “helicon” system, a high-field-side launched lower hybrid system, and NBI upgrades. With these, operation at higher magnetic field (from 2.17 to 2.5 T), larger plasma volume and stronger shaping, and new advanced divertors is projected to enable access to sustained operation with many key normalized parameters matching those of compact fusion pilot plant design studies, thus permitting DIII-D to assess and inform such designs.
Work supported by US DOE under DE-FC02-04ER54698.
Speaker: Christopher Holcomb (Lawrence Livermore National Laboratory) -
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Advanced Tokamak Studies in Full-Metal ASDEX Upgrade
Conventional tokamak high-confinement mode (H-mode) scenarios suffer from magnetohydrodynamic (MHD) instabilities and also depend on inductive current from the central solenoid to maintain the plasma current.
Advanced Tokamak (AT) scenarios that feature manipulated non-standard $q$-profiles not only promise to improve the stability and confinement of the discharge by eliminating some of the most common resistive MHD instabilities, such as sawteeth, but also allow to extend the pulse length by increasing the core bootstrap current density $j_\mathrm{bs} \sim q \nabla p $.This contribution presents an overview over recent Advanced Tokamak (AT) studies undertaken in the full-metal ASDEX Upgrade (AUG) tokamak in the context of paving a way to an envisaged steady-state EU-DEMO scenario [1] ($q_{95} \approx 4.5 ; \beta_\mathrm{N} \approx 3.5 ; H_{98}(y,2) \approx 1.2$).
Designing such larger next-generation devices requires a thorough theoretical understanding which allows to credibly extrapolate existing models from present-day experiments.To this end, the experimental thrust at AUG aims at providing the means to verify state of the art models for equilibrium, stability and transport physics in AT plasmas.
The parameter space covered so far experimentally can be broadly divided as follows: \
1) $q_{95} \approx 5$ plasmas with conventional off-axis co-current drive (co-CD) for $q_\mathrm{min} \approx 1.1$ and sustained $\beta_\mathrm{N,max} \approx 2.7$ [2],
2) $q_{95} \approx 5$ plasmas with on-axis co-CD for maximum current drive efficiency and anomalous central flux diffusion ("flux pumping") to maintain $q(0)$ clamped at 1, resulting in plasmas with transient $\beta_\mathrm{N,max} \approx 3.7$ [3] and sustained $\beta_\mathrm{N} \approx 3.1$ at higher shaping, and
3) $q_{95} \approx 4$ plasmas with central electron-cyclotron counter-CD to maintain $q_\mathrm{min}$ of up to $1.6$ [4].Careful tailoring of the plasma edge properties allowed excellent confinement, transiently as far up as $H_{98}(y,2) \approx 1.6$.
This approach has enabled various modelling advances, such as confirming the fidelity of equilibrium and current drive models or verifying ideal stability codes.
Furthermore, it allowed investigations into the physics behind flux pumping, and to identify gaps in existing fast transport solvers related to turbulence stabilisation through interaction with energetic particles.This contribution will report on the experimental and modelling results outlined above.
References:
[1] H. Zohm et al 2017 Nucl. Fusion 57 086002
[2] A. Bock et al 2017 Nucl. Fusion 57 126041
[3] A. Burckhart et al, IAEA FEC 2020, Experimental Evidence of Magnetic
Flux Pumping at AUG
[4] J. Stober et al 2020 Plasma Phys. Control. Fusion 62 024012Speaker: Alexander Bock (Max Planck Institute for Plasma Physics)
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15:25
Coffee
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PWI sessionConvener: Sebastijan Brezinsek (Forschungszentrum Jülich)
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Control of fuel particle recycling using the hot wall on all-metal plasma facing wall in QUEST
QUEST (Q-shu University experiments with Steady-state Spherical Tokamak) [1] is a medium sized spherical tokamak (ST) (R=0.64m a=0.4m, BT<0.25T @R=0.6m) in Kyushu University. The heating sources for plasma are two RFs which are capable of operation with 28GHz, 350kW, a few second and 8.2 GHz 50kW, CW. Plasma facing walls (PFWs) are mainly composed of stainless steel type 316L and stainless steel type 316L coated with atmospheric plasma sprayed tungsten (APS-W). QUEST has been equipped with a hot wall coated with APS-W [2], which is capable of controlling its temperature locally even during the discharge through the combination use of heater and water cooling up to 723K that is expected to be the operation temperature in Japanese DEMO. QUEST could demonstrate 6 h discharges [3] with the hot wall in a limiter configuration with low power range (<40kW).
Results from QUEST show that the wall temperature considerably influences fuel recycling property including wall saturation, leading to density runaway and termination of discharge. Higher wall temperatures have led to shorter discharges in the range of 393K(2h15min)-673K(40min) in the same condition of plasma and PFWs. This tendency can be explained as a result of the plasma-induced deposition layer that modifies surface recombination accompanied by a transport barrier for fuel hydrogen [4], locating at the boundary between the deposition layer with substrate. The surface recombination coefficient of the deposition layer mainly composed of plasma-sputtered carbon is increasing with its temperature and consequently, active surface recombination gives rise to wall saturation in early stage of the discharge. This must be resolved to obtain SSO.
In 2021, QUEST successfully demonstrated long duration discharges with wall temperature of ~673K. Unfortunately, a number of heaters of the hot wall was not working during the discharge, because they have made a damage for long time operation since 2015, but almost half of the hot wall reached to up to 673K and the rest of them was around 573K or less. In this situation, we could try to regulate the hot wall temperature and the local hot wall temperature kept in 673K was cooled down by 90 K on two occasions in the discharge during wall saturation and the global recycling rate was recovered down to less than unity each occasion, and consequently, the pulse duration extended from 40 min to 230 min. This is a good example of real time controllability of fuel recycling with local PFW temperature control.
[1] Hanada, K., et al., (2017) Nuclear Fusion, 57(12), [126061]. [2] Hasegawa, M., et al., (2018) Fusion Engineering and Design, 129, 202-206. [3] Hasegawa, M., et al., (2021) Plasma and Fusion research, 16, 2402034. [4] Hanada, K., et al., (2019) Nuclear Fusion, 59(7), [076007]. [5] Hanada, K., et al., (2019) Nuclear Materials and Energy, 19, 544-549. [6] Hanada, K., et al., (2021) Nuclear Materials and Energy, 27, [101013].Speaker: Kazuaki HANADA -
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Steady state control of fuel recycling for long pulse discharges in EAST tokamak with full metal first wall
Fuel recycling is one of the key issues for long pulse operation for tokamaks. During long pulse operation in tokamaks, the accumulation of fuel particles on the first wall leads to a decreasing of wall pumping capability, and eventually the wall changes to outgassing from pumping due to the accumulation of fuel retention and the increasing of surface temperature. This enhances fuel recycling and may leads to uncontrollable plasma density, and further plasma confinement would be deteriorated, or even plasmas would be disrupted. To some extent, the capability of long pulse operation of a tokamak depends on fuel recycling control.
Various methods have been studied in EAST tokamak for recycling control. Lithium coating and real-time lithium injection are very effective to control recycling, and the wall pumping rate is almost proportional to the Li-II emission intensity. The Divertor Dα emission is proportional to divertor neutral pressure in a wide range of auxiliary heating power of 0 – 11 MW, and control of divertor neutral pressure by divertor cryopumps is very important for recycling control. Moreover, a short distance between strike point on divertor plate and pumping slot could improve the particle exhaust. The global recycling coefficient was decreased notably with Resonant Magnetic Perturbation (RMP) coil current, but the plasma stored energy was also decreased.
1056 s long pulse discharge was achieved in EAST tokamak, with heating power of 1.5 MW, plasma density of 1.8×1019 m-3 and Double Null (DN) configuration. Both divertor cryopumps and real-time lithium powder injection was employed. Plasma density was controlled very well during the whole discharge. Particle balance analysis shows that the pumping rate of divertor cryopumps was ~ 2.0×1020 D/s. The wall pumping rate became almost zero since 60 s, and it was kept at zero until 800 s. After that the wall outgassing was gradually increased but was limited to a low value of ~0.4×1020 D/s. The H/(H+D) ratio from spectrometry diagnostics increases from ~700 seconds. This means that the outgassing from some part of the first wall was enhanced since ~ 800 seconds, due to the gradual increase of the first wall temperature. However, the total outgassing rate was limited due to the continuous real-time injection of lithium powder, and the divertor cryopumps provided ~ 5 times higher exhausting rate than the wall outgassing. Therefore, the fuel recycling was well controlled during the whole discharge of 1056 s, the global recycling coefficient was kept in the range of 0.95 - 0.97. The successful control on fuel recycling during long pulse discharge of over thousand seconds provides valuable references on the steady state operation of future fusion reactors, such as ITER.[1] Maingi, R., et al., J. Fusion Energy 39 (2020) 429–435
[2] Yaowei Yu, et al., Nucl. Fusion 59 (2019) 126036
[3] Yaowei Yu, et al., Phys. Scr. T170 (2017) 014070Speaker: Dr Yaowei Yu (Institute of Plasma Physics, Chinese Academy of Sciences) -
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Development of advanced vacuum and wall conditioning technologies for extending plasma pulse duration for EAST
Effective control of fuel recycling and impurity is very key for achievement of long pulse and high-performance plasmas. High recycling and impurity concentration in the plasma would result in usually degradation of plasma confinement, and uncontrollable plasma density and disruptions. Some advanced vacuum and wall conditioning technologies, have been developed and widely used in EAST to effectively control fuel and impurity particle for extending plasma pulse duration.
In order to enhance particle exhaust rate for improving recycling control capability during long-pulse plasma operation, the new type ITER-like W–Cu divertor with larger conductance was successfully applied in EAST to replace the previous graphite divertor with the higher hydrogen retention characteristic. Based on the new divertor and upgraded pumping system, the effective pumping speed for D2 of the EAST lower divertor increased about 2 times. Several techniques of surface conditionings such as baking, discharge cleaning, silicon/lithium(Li) coatings, and liquid Li, have all been attempted to further improve wall condition[1, 2]. Compared to traditional wall conditionings, it is noted that evaporated Li coating assisted by He-ICRF was testified as the most effective way to suppress impurities, reduce recycling and H/(H+D) ratio to stabilize plasma edge[1, 3]. Specifically, the high-Z tungsten core impurity concentration was maintained between 3 ppm–15 ppm during long H-mode plasmas, which is possibly due to the effect of Li film physical isolation and Li vapor shielding between the W substrate and plasma. These results confirm Li coating served as a sacrificial protective layer and can reduce wall material surface erosion[4], which will produce material of 103-105kg/year in pilot plants. In order to refresh Li coated film with the short lifetime of ~300s to continuously capture particle during long pulse discharge[5], real-time Li injection was successfully applied, which could further reduce the recycling by 30% and kept low tungsten impurity due to decreased tungsten sputtering rate resulting from the reduced divertor electron temperature[6, 7].
By using these advanced vacuum and Li wall conditioning technologies, a record plasma of ~1056s pulse duration with a controlled plasma density of 1.8×1019 m-3, the low H/(H+D) ratio to <7%, goal recycling coefficient <1 and core tungsten impurity concentration~5.6×10-5 was successfully achieved in EAST, serving as important references for stand-steady plasma operation for ITER and future fusion devices.[1] G. Z. Zuo, J. S. Hu, S. Zhen, et al., Plasma Phys. Control. Fusion 54, 015014 (2012).
[2] G. Z. Zuo, J. S. Hu, R. Maingi, et al., Nucl. Fusion 59, 016009 (2019).
[3] G. Z. Zuo, J. S. Hu, Y. W. Yu, et al., Fusion Eng. Des. 131, 41 (2018).
[4] Z. L. Tang, G. Z. Zuo, C. L. Li, et al., J. Nucl. Mater. 555 (2021).
[5] C. L. Li, G. Z. Zuo, R. Maingi, et al., Plasma Phys. Control. Fusion 63, 015001 (2021).
[6] W. Xu, J. S. Hu, R. Maingi, et al., Fusion Eng. Des. 137, 202 (2018).
[7] G. Z. Zuo, J. S. Hu, R. Maingi, et al., Nucl. Fusion 57, 046017 (2017).Speaker: Mr Guizhong Zuo (Institute of plasma physics, HIPS, Chinese academy of Sciences)
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LPO sessionConvener: Joerg Stober (IPP Garching)
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DIII-D and EAST research towards long-pulse high-performance tokamak operation
Coordinated experiments on DIII-D and EAST are developing the physics basis of fully non-inductive, high poloidal-beta (βP) plasmas for application to steady-state high performance operating scenarios in ITER and Fusion Pilot Plants (FPPs). By optimizing at low plasma current and high plasma pressure, high-βP operation reduces disruption risks and requirements on external current drive, while improving the energy confinement quality through Shafranov shift suppression of turbulence (α stabilization). The robust α stabilization mechanism scales favorably with pressure, and does not require external momentum injection, nor a carefully tailored current density profile.
On DIII-D, high βP and high density have enabled fully noninductive operation with highly self-organized plasma profiles exhibiting robustness to external perturbations and excellent confinement quality (H98≥1.5). The values of βN≥3.5, H98≥1.5, q95~7 achieved simultaneously on DIII-D, match the normalized performance of a Q~17 compact fusion pilot plant design point [1]. Recent experiments have achieved world-leading results in core-edge integration research, demonstrating high core performance simultaneous with full detachment and small/no ELMs [2].
Experiments on EAST have made progress in extending performance of long pulse H-mode using RF-only heating and current drive [J. Huang, this meeting]. Even though these latest experiments achieved βP and βN values comparable to those that in DIII-D result in the formation of a large radius ITB (βP>2 and βN~2) values, an ITB is only visible in the Te profile, at ~0.3. Transport analysis reveals that the EAST plasmas are limited by ITG turbulence, despite the low ion temperature gradients. Adding heating to electrons or ions in the modeling cannot significantly increase the pressure gradient at mid-radius, unless the q-profile is modified with a higher qmin, or a deep fueling source is added. Various experimental approaches are being pursued on EAST, including early heating or broader profiles of the external current drive to create a high qmin profile. Results of upcoming experiments will be discussed.[1] R.J. Buttery et al., Nucl. Fusion 61 (2021) 046028
[2] L. Wang et al, Nature Comm. 12 (2021) 1365Supported by US DOE under DE-FC02-04ER54698 and DE-SC0010685, National Natural Science Foundation of China under 11922513, and 11775264 and National Magnetic Confinement Fusion Science Program of China under 2017YFE0301300 and 2017YFE0300404
Speaker: A. M. Garofalo (General Atomics) -
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The Advanced Tokamak Path to a Compact Fusion Pilot Plant
The Advanced Tokamak concept represents a virtuous approach for a fusion reactor, combining improved confinement and stability with reduced heat flux and disruption severity, and the potential for fully stationary “always on” steady state operation to ease engineering and stability challenges. Self-consistent, integrated 1.5D simulations project new paths to a compact fusion pilot plant based on this approach that could demonstrate net electricity production and conduct long pulse nuclear testing. The concept benefits from a combination of strong shaping, broad profiles and high beta operation, to reduce turbulent transport, raise pedestals, and raise or remove various global MHD and energetic particle drive instability limits. This leads to configurations where the plasma becomes self-driven by bootstrap currents that naturally align to the required profiles.
The physics-based approach deployed here leads to new insights and understanding of reactor optimization. Studies utilize a new integrated 1.5D core-edge approach for whole device modeling to predict plasma performance, by self-consistently applying the latest transport, pedestal, equilibrium, stability and current drive physics models to converge fully non-inductive stationary solutions without any significant free parameters. This contrasts with previous “systems code” approaches, where parameters are simply set to desired values based on plausible arguments.
Studies highlight the critical levering roles of density, toroidal field and beta in increasing fusion performance, while raising stability and enabling increased heat dissipation at higher density. The resulting increased confinement and bootstrap fraction reduces heating and current drive demands as a fraction of fusion power, and thus enables configurations with high net electricity at reduced size and current. Solutions are found with ~200MW net electricity at the 4m major radius scale and 6-7T. Heat loads are mitigatable with reasonable levels of core and divertor radiators, and good H mode access maintained. Auxiliary current drive is projected from neutral beam and ultra-high harmonic (helicon) fast wave, though other advanced current drive approaches presently being developed also have potential, and may be more desirable.
Low recirculating power and a double null configuration leads to a divertor heat flux challenge comparable to ITER, though reactor solutions may need to increase dissipation further. Strong H-mode access (factor >2 margin over the L-H transition scaling) and ITER-like heat fluxes are maintained with ~20-60% core radiation. Neutron wall loadings appear tolerable but suitable for a nuclear testing mission. The approach would benefit from high temperature superconductors, the higher fields of which increase performance margins, while their potential for demountability would facilitate a nuclear testing mission. An advanced load sharing and reactive bucking approach in the machine centerpost region provides improved mechanical stress handling.
As with all power plant concepts, the work identifies some significant plasma and technology research challenges, though the inherent advantages of the approach discussed here have the potential to offer a more stable path to a power plant in the near term, and ease challenges placed on technology components. This presentation will explain the underlying physics benefits, their experimental validation and the projection to compact solutions for a fusion pilot plant.
Speaker: Richard Buttery (General Atomics) -
13
Plans for long-pulse operation in JT-60SA
JT-60SA is a large fully superconducting new tokamak device built jointly by Europe and Japan [1]. The tokamak was fully assembled in March 2020, and the integrated commissioning of the tokamak is on-going. The mission of JT-60SA is to contribute to the early realization of fusion energy by addressing key physics issues for ITER and DEMO. Especially, development of fully non-inductive steady-state high $\beta_N$ operations above the no-wall ideal MHD stability limits, for long time exceeding the current diffusion time is an important target of JT-60SA.
The machine enhancement of JT-60SA takes a staged approach. There are three major phases in JT-60SA: the initial research phase, the integrated research phase, and the extended research phase. In the initial research phase, power of 33 MW ($P_{\rm NNB}$/ $P_{\rm PNB}$/ $P_{\rm EC}$=10/ 20/ 3 MW) and plasma current up to 5.5 MA are available. High power and high plasma current operation will be demonstrated. As an inertial cooled lower divertor with CFC tiles will be used in this phase, the pulse length with high heating power experiment will be limited to about 5 seconds. In the integrated research phase, power of EC will be upgraded to 7 MW and the divertor will be upgraded by an actively cooled divertor. Those enhancement will enable us to try steady-state long pulse operation at a divertor heat load of 10 MW/m$^2$. We also have a plan to replace the carbon divertor by the tungsten-coated divertor to develop long pulse operations compatible with the metal wall environment. In the extended research phase, further enhancement of heating and current drive is envisioned.
Significant modelling activities have been conducted to predict the plasma performance in the long-pulse operation. A strong predict first activity has been carried out with integrated modelling codes such as TOPICS and CRONOS. It is shown that the expected values for $\beta_N$, H98y2 and fBS, as described from 0-D studies in the JT-60SA Research Plan [2], can be attained in long-pulse operation. To prepare for hybrid scenarios, the safety factor profile control using off-axis ECCD during Ip ramp-up phase is studied and it is found that q > 1 can be maintained if 2.2 MW of ECCD is applied at $\rho$ ~ 0.3. To realize a high $\beta_N$ long pulse operation with high radiation fraction, multi-impurity seeding scenario is studied and it is found that high radiation in the divertor region without increased radiation in the core region can be envisioned with a mixed Ar and Ne seeding.[1] P. Barabaschi et al., Nucl. Fusion 59 (2019) 112005.
[2] JT-60SA Research Plan - Version 4.0, Sept. 2018, http://www.jt60sa.org/pdfs/JT-60SA_Res_Plan.pdfSpeaker: Takuma Wakatsuki (QST)
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11
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10:35
Coffee
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LPO sessionConvener: Joerg Stober (IPP Garching)
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14
The Spherical Tokamak for Energy Production (STEP): a Steady-State Fusion Reactor
The UK-based STEP programme aims to develop by 2040 a prototype reactor based on the spherical tokamak (ST) concept, thereby establishing a basis for developing commercial electricity production from fusion [1]. The compact design restricts the possible inductive flux, hence the flat-top plasma current will be entirely non-inductive, enabling long-pulse operation. External current drive will be delivered using electron cyclotron and electron Bernstein waves. Modelling using the transport code JETTO, with input from the systems code PROCESS, indicates that the time required to ramp up the current will be around 1000s or longer, in part to avoid large back-EMFs. To demonstrate the viability of an ST-based commercial reactor, the flat-top phase will ultimately need to be much longer than the ramp-up, while the energy confinement time will be a few seconds. STEP will therefore need to be a steady-state device, with a fuel cycle and control system that keep the plasma close to a target scenario. The ST concept makes it possible to maximise fusion power and bootstrap fraction in a compact device at relatively low toroidal field by allowing operation at high normalised pressure and elongation, but it also poses unique challenges. Double null operation will be used, with divertors that are resilient to losses of up-down symmetry. Operating in a highly self-organised, high beta scenario with large radiation and bootstrap fractions will require novel control techniques. The prediction of confinement is particularly challenging. For a high beta ST such as STEP, presently-available scaling laws and reduced transport models are well outside their domain of experimental validation. Moreover, parameter dependencies differing from those in scaling laws derived for conventional tokamaks have been observed in present day STs, and reduced models need to be modified due to the electromagnetic nature of the turbulence expected to dominate. Linear gyrokinetic (GK) modelling shows that the turbulence in reference flat-top plasmas is dominated by micro-tearing modes (MTMs) and kinetic ballooning modes (KBMs). Diamagnetic flow shear is stabilising for KBMs and other modes at high wave number k, and the turbulence is likely to be dominated by MTMs at low k: this makes the non-linear GK modelling very challenging. Actuators to control the transport are being investigated to seek routes to optimised confinement. Reduced models capturing the magnetic flutter-driven electron heat transport are being tested using the integrated modelling suite JINTRAC. Results so far support assumptions made for the simpler Bohm-gyro-Bohm model used for the concept evaluation. We have identified a preferred scenario based on the exploration of several prototype concepts with different plasma and technology assumptions. In this scenario fusion alpha-particle-driven toroidal Alfvén eigenmodes are predicted to be suppressed due to high beta. I will give an overview of STEP and discuss the physics design of the preferred scenario. I will present the key challenges and assumptions leading to the scenario choice and discuss the modelling framework being used to reduce uncertainties. This work gives confidence that a compact fusion reactor will be feasible.
[1] https://step.ukaea.uk/
Speaker: Ken McClements (United Kingdom Atomic Energy Authority) -
15
Experimental study of core MHD events in thousand-second improved confinement plasma in EAST
Recently, stationary plasma with a world-record pulse length of 1056 second was obtained, where a stable internal transport barrier (ITB) is present in electron temperature channel. The core magneto-hydrodynamics (MHD) events with m/n=1/1 or m/n=3/2, m is the poloidal mode number and n is the torodial mode number, have been observed near e-ITB region. The time evolution of frequency and the 2-D structures of these modes are studied by combination of soft X-ray (SXR) imaging and electron cyclotron emission (ECE) diagnostics. The m/n=1/1 mode exhibited with a feature of frequency chirping down in time with a chirping rate corresponding to the rate of electron diamagnetism drift frequency change. A twisted pattern is reconstructed by SXR tomography of m/n=1/1. The m/n=3/2 has a smaller frequency comparison to m/n=/1, and carries an m/n=3/2 island with detectable size. The destabilization of core MHD modes are due to a combination effects of strong central heating by electron cyclotron resonance heating (ECRH) and lower hybrid current drive (LHCD). It is found the m/n=3/2 mode is dominate before t < 23 second, and m/n=1/1 becomes dominate later. Transitions between m/n=3/2 and m/n=1/1 is found in the entire discharge after t=23 second. Self-regulation system resulted from multiscale interaction between core MHD instabilities and electron temperature gradient induced turbulence is a contributing mechanism for sustaining the steady state long pulse high confinement regime. A negative current is generated in the magnetic axis with m/n=1/1, which anomalously broad the core current profile. The transition between m/n=1/1 and m/n=3/2 is due to the mode coupling due to forced magnetic reconnection induced by m/n=1/1 mode. The interaction between MHD modes and fast electrons and the actively control of those modes are discussed.
Speaker: Liqing Xu (ASIPP)
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14
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12:20
Lunch
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LPO Control sessionConvener: Tomohiro Morisaki (National Institute for Fusion Science)
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16
Establishing fusion reactor control scenarios based on information from a reduced set of nuclear-compatible diagnostics
Existing magnetically confined plasma devices benefit from an extensive array of diagnostics, commensurate with the R&D function of these plasma devices. While increased diagnostic coverage, and access to information relevant to the plasma, first-wall components, and plasma-material interactions is always desired, the harsh nuclear environment of future fusion reactors is more likely to result in the situation where operators (AI or human)1,2 will have access to less information, compared to what is currently measurable in existing pathway research devices. Future fusion reactors will be industrial, energy-production devices, and operate in a functionally reduced area of parameter space, which may be consistent with the limited information that will be available. Stated as a question: what is the least/critical amount of information that industrial fusion reactors will need to operate robustly and safely? And consequently, what nuclear compatible diagnostic systems are needed to make the measurements that will provide that information?
Answering these questions requires demonstration of a burning plasma, which is within the goals of the ITER Research Program3 and devices planned by private industry. While ITER aspires to achieve a sustained Q=10 fusion plasma, it will also have a diagnostic set that is similar to that found on R&D devices. Moreover, diagnostic technology will evolve as a result of experience in the nuclear environment of ITER. And data analysis techniques (ML, integrated modeling, etc.) will redefine what information can be derived from the available set of physical measurements. A successful US fusion pilot plant (FPP) design will need to incorporate knowledge gained from ITER in these areas: 1) nuclear compatible diagnostic designs, 2) integrated modeling that can extract critical, indirect information from direct measurements using those diagnostics, and 3) reactor control scenarios that utilize that information to operate robustly and safely in a specific configuration for fusion energy production.
This presentation will describe the efforts at ORNL that endeavor to address the research that is needed in these areas for long-pulse fusion plasma devices.1 A. Katwala, “DeepMind has Trained an AI to Control Nuclear Fusion”, WIRED, Feb 16, 2022. https://www.wired.com/story/deepmind-ai-nuclear-fusion/
2 J. Degrave, F. Felici, J. Buchli, et al. “Magnetic control of tokamak plasmas through deep reinforcement learning.” Nature 602, 414–419 (2022). https://doi.org/10.1038/s41586-021-04301-9
3 ITER Organization, “ITER Research Plan within the Staged Approach (Level III – Provisional Version),” ITR-18-003, Sep 17, 2018.This work was supported by the U.S. D.O.E contract DE-AC05-00OR22725.
Speaker: Theodore Biewer (Oak Ridge National Lab) -
17
Real-time plasma control of fully non-inductive operation in EAST 1056s long pulse discharge
Feedback controlled fully non-inductive plasma discharge have been sustained in EAST long-pulse operation up to 1056s with a new world record of injected–extracted energy exceeding 1.7 GJ. This steady-state real-time plasma control requires integrated accurate control of plasma equilibrium, current and loop voltage [1]. An improvement of the plasma position and shape control within a few millimeters range, together with zero loop voltage control by the injected LHW power. The fiber optic current sensors (FOCS), based on the Faraday Effect, with no signal drift, are firstly used and provide higher precision current measurement compared with Rogowski coil in plasma current control [2]. Low zero drift integrators are applied for magnetic measurements used to reconstruct and control the plasma position and shape [3]. Small drifts were observed during the dry-run operation and corrected in EAST plasma control system (PCS) for long-pulse operation. Magnetic measurement signals selection and fitting uncertainty in real-time equilibrium reconstruction are based on error analysis with 500s dry-run operation. With delicately optimizing loop voltage PID controller and 4.6Ghz LHW feedforward power, loop voltage is controlled within 0.5×10^-5 V. Advanced tools, experimental results and analysis in EAST long-pulse oriented plasma control provide important experience and reference for long-pulse operation on fusion devices.
References
[1] Yuan, Q. P., et al. "New Control Abilities on EAST PCS for Steady-State Operation." IEEE Transactions on Plasma Science (2018):1-5.
[2] Xue, M. M., et al. "Fiber-optic current sensor for plasma current on experimental advanced superconducting tokamak." Fusion Engineering and Design 140 (2019):11-15.
[3] Wang, Y., et al. "A New Analog Integrator for Magnetic Diagnostics on EAST." IEEE Transactions on Nuclear Science (2019):1-1.Speaker: YAO HUANG (ASIPP) -
18
Control and protection challenges in fully metallic tokamak WEST for long pulse operation
WEST is a tungsten tokamak designed for long pulse operation. It is now fully equipped with actively cooled components especially, the lower divertor is made of 456 ITER like Plasma Facing Units (PFU) that are being qualified in a tokamak environment. The main missions of WEST address high fluence plasma divertor exposure and the demonstration of long pulse H-mode in a full tungsten environment. In particular, thanks to the actively cooled plasma facing components (PFC), the fluence of a single ITER discharge can be obtained in a few long WEST discharges while several months of operation would be necessary on tokamaks equipped with inertial PFCs. Achieving long duration and high performance plasma discharges while ensuring the protection of the machine requires specific features such as dealing with unexpected events that may occur at any time. As a result, the WEST Plasma Control System (PCS) addresses the plasma discharge as a set of sequences also called segments [Nouailletas 2019] each of them being triggered by an event. For example, several segments have been designed to deal with the discharge soft landing to address runaway mitigation, flux swing flux limit etc. In addition, each controlled parameter can vary in a preset range defined by an envelope that allows an acceptable deviation from the preset value without generating an event resulting in triggering a new segment. The development of plasma scenarios is also relying on specific controls. The simultaneous controls of the loop voltage and the plasma current, using respectively the voltage applied to the central solenoid and the power of the lower hybrid current drive system is available and used for long pulse operation. An important mission of WEST consists in testing the ITER-like PFUs submitted to a constant and controlled heat load. Recently the real time evaluation of this load has been implemented permitting its control through the auxiliary heating power. In parallel to the performance and duration, the machine protection must be guaranteed at any time. This function is ensured by an extensive set of diagnostics, coupled with advanced controls and cross diagnostic functions. Among them, the infrared system that monitors all critical in-vessel components plays a central role. Indeed, the accurate conversion to surface temperature from the infrared radiance maps is complex due to various phenomena that must be considered in a metallic environment such as reflections, and low and varying emissivity. Finally, even more advanced control functions based on artificial intelligence are under development. They offer a greater adaptability to different scenarios, while being more robust with respect to the uncertainties on the plasma characteristics and the non-linearity of involved dynamics. At present, they are tested on a simulator of the WEST tokamak.
[Nouailletas2019] R. Nouailletas et al. Fusion Engineering and Design 146 part A (2019)
https://doi.org/10.1016/j.fusengdes.2019.01.139Speaker: Philippe MOREAU (CEA)
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16
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15:25
Coffee
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LPO Control sessionConvener: Hans-Stephan Bosch (Max-Planck-Institute for Plasma Physics)
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19
Model based formation of Advanced Tokamak discharges
A model has been developed in the transport code ASTRA, which is capable of simulating advanced tokamak discharges, using the density and actuator setup as inputs. The model uses a reduced Gyro-Bohm based core transport, which does include simplified ITG and TEM mode contributions, to achieve a run time of only a few minutes for a full discharge. Edge transport is included via the use of a recently developed scaling law. Also included is the L-H transition, which is triggered based on the heating power at the separatrix. The model does include a set of free parameters, which have been adjusted using a set of reference discharges. These parameters are consistent between discharges, as long as no major scenario changes are done. For a sufficiently different scenario, they would have to be re-evaluated.
With this setup it is possible to quickly test a large amount of possible actuator changes to a reference discharge, allowing for a large part of scenario development to be done through modelling only. This is especially relevant for early heating scenarios, where fully experimental scenario design can cost a lot of discharges. This approach allows for an optimal entry to a desired q-profile early in the discharge, without passing through an unfavorable regime, which is important for present devices due to short pulses but also for a reactor due to its long current diffusion time.
The model was used to design an early-heating discharge at AUG, which was run successfully, showing its viability to be used as a tool for scenario design. In this discharge, co-ECCD was used to achieve a $q_{95}$ of about 5.2. Stable operation for this scenario was possible for values of $\beta_n$ between 2 and 3, the stability limit was found at $\beta_{n}\sim 3.2$. Results of this scenario will be shown.
The model was tested in multiple different scenarios at different plasma currents and $\beta_n$ for both co- and counter-current ECCD operation. A scenario with anomalous flux redistribution (flux pumping) was also investigated.A counter-ECCD scenario with a higher current than the validation scenario, reaching $q_{95}$ of $\sim 4.1$ has been investigated. Due to the non-availability of one of the two current drive NBIs at AUG last campaign, this scenario needed to be adapted to run with a different current drive setup than in previous campaigns. Using the model, it was shown that the missing NBI current drive can be substituted by some of the remaining systems and the scenario was successfully run with the changed current drive setup.
An optimizer, built around the RAPTOR fast core transport solver [Sauter, Plasma Phys. Control. Fusion 54 025002] was used to propose changes to the actuators for this scenario to optimize the q-profile. The goal was to improve NTM stability, performance, and reach a stationary elevated q profile early in the discharge. The effect of these changes were checked in ASTRA, before running them on the experiment. Results will be presented.
Speaker: Raphael Schramm (Max Planck Institute for Plasma Physics) -
20
Avoidance control of high-density collapse based on data-driven prediction in Large Helical Device
Avoidance of radiative collapse in high-density plasma has been attempted in the Large Helical Device (LHD) with a real-time control system based on a data-driven predictor model. The predictor model has been developed based on machine-learning techniques and high-density experiment data in LHD.
In stellarator-heliotron plasma, radiative collapse is one of the most critical issues that limit the performance of plasmas, while the stable high-density operation is an advantage of a helical system over a tokamak. In our previous research, low-Z impurities and edge plasma temperature were extracted as the features of radiative collapse with a machine-learning model and sparse modeling. Using these features, the possibility of the occurrence of the radiative collapse was quantified as the collapse likelihood [a].
A single-board computer, Raspberry Pi 4, has been used as a controller that calculates the collapse likelihood in real-time and alarms when the likelihood exceeds the threshold value, which means the plasma is approaching the collapse. The boost ECH is injected and gas puff fueling is turned off while the alarm signal is issued.
The control system has been applied to the density ramp-up experiment in LHD [b]. In the ramp-up in the early phase of discharge, the predictor detected a radiative collapse. Without control, the plasma was shut down immediately after the detection. When the control system is employed, the boost ECH and turning gas puff off were triggered and the collapse was avoided successfully. After avoidance, the plasma density kept developing moderately. In the latter phase of the discharge with control, collapses were avoided by turning gas puff on/off and the electron density was developed above $1.2\times10^{20} \mathrm{m}^{-3}$. It has been also attempted to avoid radiative collapse only with boost ECH by tuning the injection setting.
The authors are grateful to the LHD experiment group for the excellent support of this work. This work is supported by the National Institute for Fusion Science grant administrative budgets NIFS21KLPP068, and JSPS KAKENHI Grant Numbers JP19J20641 and JP19H05498.
References
[a] T. Yokoyama, et al. Journal of Fusion Energy 39, 500–511 (2020).
[b] T. Yokoyama, et al. Plasma and Fusion Research 17, 2402042 (2022).Speaker: Dr Tatsuya Yokoyama (Naka Institute, National Institutes for Quantum and Radiological Science and Technology) -
21
Real-time feedback and plasma controls for steady-state plasma operation
Demonstration of long-pulse plasma discharge is one of the critical issues in making fusion reactors true. In the Large Helical Device (LHD), we realized the duration time of 48 min with the electron density of 1.2e19 m-3, the plasma temperature of 2 keV, and the heating power of 1.2 MW using the ICRF and ECRF waves in hydrogen minority helium plasmas He(H). Several feedback systems kept the plasma parameters constant, and we could successfully control fueling without perturbation for electron density. For auxiliary heating, injection power was real-time controlled with an FPGA circuit, and boost injection for heating power could mitigate gradual density rising associated with outgassing and rapid density rising in unintended impurity contamination. The critical time for real-time feedback was approximately energy confinement time ( < 100 ms). If there was no additional heating support during energy confinement time, plasmas were terminated just after events.
Controlling the concentration of hydrogen particles for ICRF heating during long-pulse plasma duration is one of the essential tasks for keeping heating efficiency because single-pass absorption for minority heating was strongly associated with minority concentration. The optimized single-pass absorption for minority heating was 3 ~ 5 % around ICRF resonances, and accurate minority concentration controls are required during long-pulse plasma duration. In hydrogen minority deuterium or helium plasmas, D(H) and He(H), we have studied particle confinement time with superimposed hydrogen gas puffing with the same frequency and injection during long-pulse plasma. The time evolution of particle confinement time (p) gradually increased in a few ten seconds, and then two kinds of p were observed. The initial expectation of the time evolution of p was gradually increased, and then finally, we could get the single-particle p for the primary gas fueling. However, this experiment phenomenon was different from the initial expectation, and it seemed to be one of the long-pulse phenomena caused by wall recycling and particle fueling.
In this paper, we show the heating control associated with the plasma parameters, the recovery operation from the unintended impurity contamination, and the time evolution of particle confinement time during long-pulse plasma duration in D(H) and He(H) plasmas. Finally, we suggest a minority concentration control scheme during ICRF heating in long-pulse plasma duration.Speaker: Dr Hiroshi Kasahara (NIFS) -
22
Development of core plasma density feedback control necessary using fuelling pellets
In long pulse experiments on ITER and DEMO, pellet injection will need to be used to fuel the core plasma and will likely be the sole actuator available for core density control. The reason is that gas injection will be largely ineffective due to its limited penetration depth towards the core. However, considering the expected 10% lost pellet rate during injection, feedback control will be vital to compensate for the lost pellets and retain or regain preferable density profiles. Moreover, the size of pellets plays a crucial role on the design of the feedback control strategy. This work outlines several challenges associated with pellet (feedback) control and shows some results from the first model predictive controller designed to do profile control with multiple pellet injectors.
Speaker: Matthijs van Berkel (DIFFER)
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19
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PWI sessionConvener: Kazuaki HANADA
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23
Implications of net erosion and redeposition of solid-surface plasma facing material in long-pulse fusion devices
It is estimated that long-pulse fusion devices may experience rates of net erosion and deposition of solid PFC (Plasma Facing Component) material of 10^3 – 10^5 kg/year, whatever the material used [1]. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘UFOs’ and unsafe dust levels. The potential implications appear to be no less serious than for plasma contact with the divertor target, i.e., a dust explosion or major UFO-disruption could be as damaging for an actively-cooled DT tokamak as target failure. Therefore, it is necessary to manage material deposits to reduce operational risk. This situation appears to require a fundamental paradigm shift regarding meeting the challenge of taming the plasma-material interface; in that any acceptable solid PFC material will in effect be flow-through, like liquid-metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable like car brake pads. The implications for such a paradigm shift and near-term research needs will be discussed.
The management of eroded material migrating within fusion devices is a well-known issue. A critical open issue is the formation of large/thick redeposited material in the divertor region, colloquially called slag. The stability of the slag layers can lead to the aforementioned ‘UFOs’ and/or dust. In JET-ILW and ITER, the use of a high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls presents unique challenges with regards to both slag formation and its stability. Furthermore, future reactors like the US ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding of the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc.) used as in situ replenishable, relatively thin (of order mm’s) claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing risk of degrading the confined plasma. In situ removal of these layers, both metallic and non-metallic, e.g., using strike-point sweeping and/or chemical ‘scavenger’ methods, requires increased research since understanding and predictive capability for the formation and stability of these thick layers is almost completely lacking.
Separately, wall conditioning has proven essential for achieving high performance. For long-pulse fusion devices, standard methods appear unworkable, but recently powder droppers injecting low-Z material ~continuously into discharges have been quite effective and may be usable in long-pulse devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers also provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current magnetic confinement devices, independent of whether their protection tiles use low-Z or high-Z material.
*Work supported in part by the US DOE under contracts DE-AC05-00OR22725,DE-AC02-09CH11466,DE-FC02-04ER54698.
[1] P.C. Stangeby, E.A. Unterberg, et al. (2022) Plasma Phys. Control. Fusion, Part A: 64 055018 & Part B: 64 055003.Speaker: Ezekial Unterberg (Oak Ridge National Laboratory) -
24
Preparation of long pulse divertor operation at Wendelstein 7-X
A fusion reactor based on a stellarator design has the advantage of easier access to long pulse scenarios. In fact, one of the main goals of Wendelstein 7-X (W7-X), the largest advanced stellarator in the world, is to demonstrate the steady-state capabilities of the stellarator line. Therefore, in the recent campaign, a number of experiments were performed in order to prepare long pulse operations, addressing issues like the development of stable detachment, control of the heat and particle exhaust, and the influence of leading edges on plasma performance. The heat and particle exhaust in W7-X is realized with help of an island divertor, which utilizes large magnetic islands at the plasma boundary. This concept shows very efficient heat flux spreading and favorable scaling with input power.
A highlight of the recent campaign was a robust detachment scenario through either intrinsic carbon or seeded impurities (Ne, N2). Detachment allowed removing of a large fraction of power loads due to direct contact of divertor target plates with the plasma while reaching neutral pressures at the pumping gap entrance yielded the particle removal rate close to the values required for stable density control in steady-state operation. Before the onset of detachment, the values of the downstream electron density are significantly higher ($1.2-1.4\cdot10^{20}$ $\mathrm{m}^{-3}$) than electron densities near the separatrix ($n_{\mathrm{e,sep}}$ is in the range $4-6\cdot 10^{19}$ $\mathrm{m}^{-3}$). This difference between upstream and downstream density indicates that the divertor operates in the high recycling regime. This detachment regime is characterized by low impurity concentration ($Z_{\mathrm{eff}}= 1.5$) in the plasma and high neutral pressure ($p_n≤ 0.1$ Pa) in the subdivertor volume. Estimates of total pumping rate for detached discharges at ca. $0.6\cdot 10^{21}$ [atoms/s] show that at these neutral pressures particle exhaust is at the level required for steady-state operation. This was sufficient for stable density control in the discharges fueled either with gas injection or pellets. Both systems will be used for long pulse operation in the upcoming experimental campaigns.
A series of experiments were performed to study the behavior of intrinsic impurities as well as seeding low and highly recycling species to enhance plasma radiation. Overall W7-X shows good impurity control in low and high density discharges with Te/Ti > 1. We have found that despite high influx of carbon into the SOL during discharges with dedicated overloading of the leading edges, the plasmas remained stable. Line-of-sight averaged Zeff stayed below 1.5 throughout the discharge and radiation increased at the plasma edge only.
The results presented in this work form a promising outlook on the overall steady-state compatibility of the detached island divertor concept in future experiments and a stellarator-based reactor.Speaker: Marcin Jakubowski (Max-Planck-Institut für Plasmaphysik) -
25
Divertor pumping for steady state operation in LHD experiments
Attempts using various particle control knobs have been made at Large Helical Device (LHD) to achieve steady-state plasmas in long pulse discharges. Divertor pumping is an important tool to control plasma density in fusion plasmas. In the divertor region, neutral particles shall be compressed and efficiently pumped out. In the LHD, the development of divertor pumping has been strongly enhanced. With the pumped helical divertor, we have achieved the low recycling state in the LHD [1]. In the conference, we will present the results of divertor pumping experiments for long pulse discharges.
Divertor pumping was applied to 40-second long pulse Electron Cyclotron Heating (ECH) discharges to assess the improvement in particle control provided by divertor pumping [2]. The results show that without divertor pumping, the electron density was not controlled only by gas puffing using the feedback signal of line-averaged electron density. Then, the plasma confinement deteriorated, finally leading to radiation collapse. On the other hand, with divertor pumping, the density was well-controlled by gas puffing using the feedback signal.
A heat transport analysis shows that divertor pumping did not affect edge electron heat conductivity, but it led to low electron heat conductivity in the core region with the formation of the electron-internal-transport-barrier. The results suggest emergence of the core-edge coupling caused by the divertor pumping.
In this study, we will also present the technical development in divertor pumps which is essential for the steady state operation.
An organic adhesive-free bonding technique, which enables outgassing-free, was developed for divertor cryo-sorption pump in LHD. The technique, which avoids the contamination of vacuum vessel by outgassing, is acceptable in future fusion devices. Also, the pumping performance is maintained during the neutron environment in deuterium experiments.
Sintering method of activated charcoal, in which pore size is optimized for the cryo-sorption pump by SPS (Spark Plasma Sintering), was developed. R&D shows that the pumping capacity with optimized activated charcoal is higher than that in commercially available activated charcoal. The technique contributes to the high performance of the vacuum pumps.
[1] G. Motojima et al., Nuclear Fusion 59 (2019), 086022.
[2] G. Motojima et al., Physica Scripta 97 (2022), 035601.Speaker: Dr Gen Motojima (National Institute for Fusion Science)
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10:15
Coffee
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PWI sessionConvener: Sebastijan Brezinsek (Forschungszentrum Jülich)
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Direct measurement of dynamic retention from plasma exposed stainless steel type 316L specimen using Fast Ejecting System of Targeted Sample (FESTA) on QUEST
The hydrogen isotopes retention and subsequent excessive desorption (dynamic retention) on the plasma facing walls (PFWs) frequently lead to density runaway and plasma termination due to R > 1, where R means recycling ratio. This issue has a significant impact on steady state operation (SSO) of fusion experimental devices. Recently, the use of metallic materials such as tungsten in PFWs on fusion experimental devices has been promoted to reduce the wall-stored hydrogen isotope. This is likely to induce the excessive desorption connect to R >1 and further investigation of dynamic retention during plasma discharge in fusion experimental devices is highly addressed. However, the measurement of the dynamic retention in fusion experimental devices is quite limited.
In this research, therefore, a device named FESTA (Fast Ejecting System of Targeted sAmple) has been developed to measure the dynamic retention of hydrogen during plasma discharge in QUEST [1]. The FESTA operation is performed by the pre-programmed way using LabVIEW. A prepared specimen is firstly set in the FESTA chamber before the plasma discharge and a FESTA arm to pick up a specimen goes into the chamber. Using the arm, the specimen is inserted into the QUEST plasma chamber to be exposed to plasma. During plasma discharge, the specimen can be extracted from plasma exposure when the arm takes it back and leaves it in the FESTA chamber. After the FESTA arm returns to the original position, the FESTA chamber is isolated by gate valves and the hydrogen desorption from the specimen is measured by a quadruple mass analyzer.
A stainless-steel type 316L specimen was serially exposed to the same plasma lasting 910 seconds three times with a fixed interval of 70 minutes at room temperature (plasma current Ip ~ 2 kA, RF power P ~ 25 kW, electron density ne ~ 1×1017 m-3, hydrogen influx Γin ~ 1.3×1017 m-2s-1). To eliminate the influence of desorption from the FESTA chamber, which is also being exposed to the plasma, a background model was developed [1]. The background model could reconstruct the hydrogen desorption from the chamber wall during the FESTA operation without a specimen. Finally, it is found that FESTA can measure the hydrogen desorption and the dynamic retention from the plasma-exposed specimen made of stainless-steel type 316L. It is observed that the dynamic retention from the specimen was increasing shot by shot, which indicates that shot history plays an important role on fuel particles, which could be explained by a diffusion-trap model with hydrogen barrier between the surface oxide layer and the material substrate developed by thermal desorption spectrum measurement of the specimen. The property of the stainless-steel type 316L around room temperature has a significant impact on fuel particle balance during long pulse operation in QUEST.
[1] Q.Yue, et al Plasma Fusion Res. 15 (2020) 240201Speaker: Mr QILIN YUE (IGSES) -
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The coupling effects of divertor configuration and drift on detachment in EAST new lower tungsten divertor for long-pulse operation
The new lower tungsten divertor with horizontal and vertical targets has already been developed and installed in EAST for high-power and long-pulse operation in a full metal wall environment. The flexible magnetic configurations allow the position of lower outer strike point located on either horizontal or vertical target with different divertor configurations. Preliminary experimental results show that the outer strike point on the horizontal target near the coroner(<2cm) can maintain easily the divertor detachment with a lower upstream separatrix density.
In this work, the coupling effects of different divertor configurations with outer strike point on the horizontal or vertical target and different E×B drift patterns on detachment in EAST are systemically investigated by the SOLPS-ITER code. The preliminary modeling results show that when the outer strike point is on the horizontal target near the corner, the vertical target can help reflect neutrals to the scrape-off layer to trap more neutrals in the divertor region. Therefore, the detachment can be achieved at a lower upstream separatrix density. In addition, E×B drift with different B×∇B directions can redistribute ions in the divertor region to either strengthen or weaken the corner effect on the detachment onset. With the B×∇B away from the X-point (rev. Bt), the E_r×B drift causes ion flows from the inner divertor area to the outer divertor area crossing the private flux region (PFR). In the outer target region, radial ion drift velocity E_θ×B pushes ions outer of PFR into the SOL. Therefore, the E×B drift with reversal Bt can strengthen the corner effect to achieve the divertor detachment with much lower separatrix density. However, when the B×∇B towards the X-point, the E×B drift will weaken the corner effect.
A more closed divertor, V-shape divertor has also been proposed as a candidate for the future EAST divertor upgrade, which will reduce the effect of different drift patterns on detachment. Compared to the present divertor configuration with the right angle, V-shape divertor configuration with both toroidal field directions directs the recycling neutrals toward the separatrix from both the SOL and PFR and focuses recycling neutrals on the V-end. This means that V-shape divertor can achieve detachment at a much lower upstream separatrix density with two different drift patterns compared to the present right angle divertor configuration in EAST.Speaker: Hang Si (Institute of Plasma Physics Chinese Academy of Sciences) -
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Effect of separatrix density on ELM instability in long-pulse H-mode plasmas on EAST
One critical issue for achieving stationary high-confinement mode (H-mode) operation of the next-step tokamak fusion reactors is to control the transient heat load induced by large-amplitude edge-localized modes (ELMs) without significant degradation of the plasma performance. Although various small/no ELM H-mode regimes have been achieved on present tokamaks, a detailed physics understanding of these regimes is not yet available. Effective and robust control of large-amplitude ELMs still awaits better understanding of the ELM characteristics for extrapolation to future fusion reactors. In this contribution, analysis of ELM instability in minute-scale long-pulse H-mode discharges on EAST and active control of large-amplitude ELMs by changing the strike point location will be presented. It has been suggested that separatrix density plays an important role in ELM instability, which is thought to be helpful to improve our understanding on the ELM instability and facilitate the long-pulse H-mode operation on EAST. In minute-scale long-pulse H-mode discharges conducted in 2017 campaign on EAST, it has been repetitively observed that the edge fuel recycling level might gradually decrease after the plasma current flattop sustained to a certain time, accomplished by the decrease in separatrix density and subsequent trigger of large-amplitude ELMs, which would induce large transient heat load on the divertor target plates and restrict the high-performance long-pulse operation. Pedestal stability analysis suggests that the decrease in separatrix density leads to an increase in the pedestal density gradient and thus higher pedestal pressure gradient and higher peak current density, which is thought to be the main reason for destabilizing medium-n peeling-ballooning modes (PBMs) and triggering the large-amplitude ELMs. By numerical scan of separatrix density, it has been found that there is a certain range of separatrix density making medium-n PBMs unstable in the pedestal region. When further increasing or decreasing the separatrix density, the medium-n PBMs might become more stable. In addition, ideal ballooning mode would become unstable at the pedestal foot region with a high separatrix density, which is thought to facilitate the achievement of small ELMs. Active control of large-amplitude ELMs by changing the strike point location has been performed with newly developed lower tungsten divertor characterized by a right-angled corner at the outer divertor in the 2021 campaign on EAST. The experimental results show that the separatrix density could be significantly enhanced as well as the scrape-off layer (SOL) density when the strike point moves to the horizontal target plate away from the corner. At the same time, large-amplitude ELMs are successfully mitigated and the plasma energy confinement performance remains unchanged. SOLPS simulation suggests that there is a high ionization source in the SOL region near outer midplane when the strike point is located on the horizontal target plate away from the corner, providing a strong fueling near separatrix and thus significantly raising separatrix density, which facilitates to trigger ballooning instabilities at the pedestal foot region and achieve the small ELM regime.
Speaker: Yifeng Wang -
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Evolution of the W cristalline structure under He irradiation: surfacic evolution and bubble formation experimental characterization
Plasma-facing materials (PFM) for next generation fusion devices like ITER will be submitted to intense fluxes of He and H isotopes (H, D and radioactive T transmutating as He). This is particularly significant for the divertor components, for which tungsten (W) is the first-choice material thanks to its low sputtering yield, low HI retention and high melting point. Plasma-wall interactions could jeopardize conservation of these key properties throughout long-pulse operation, triggering concerns for the reactor efficiency and safety: notably, incident He particles can drastically affect the surface, with the formation of dislocation loops, bubbles or W-fuzz [1]. The W surface properties plays a major role on HI retention/permeation, a crucial process for reactor efficiency and safety; it is therefore of prime importance to characterize the structural modifications triggered by He and their driving parameters in order to assess their potential impact. We will present structure evolution of various W samples submitted to He exposure, with some first T inventory results, with the main objective to improve our modelling effort and thus achieve a better evaluation of the long term inventory in W PFC of ITER through our predictive rate equation codes.
The WHIrr project [2-3] tackles the study of He bubble formation in different grades of W after exposure either to in situ conditions or laboratory devices. We have observed that He bubbles form at all irradiation temperature studies (from 60°C to 800°C), even at energies below the minimal displacement energy or low exposition fluences. Transmission Electron Microscopy (TEM) highlights a high concentration in the near surface area (0-10/25 nm deep depending on the He fluence). A strong increase in bubble average size along with a decrease in bubble density is observed as the irradiation temperature elevates over 500°C, which is the operational condition expected for the W divertor of ITER. In order to obtain insight on the kinetics of fundamental mechanisms of bubble growth in the near surface area, measurements were carried out for the first time using in situ nano science techniques: surface evolution was followed by Low Energy Electron Microscopy (LEEM), and bubble size, shape and distribution was monitored by Grazing Incidence Small Angle X-ray Scattering (GISAXS) during He irradiation at the ERSF. Size and shape evolution of He bubbles in W single crystals held at 1000°C during continuous implantation of 2 keV He ions for W(110), W(100) and W(111) showed the growth of facetted He bubbles up to 20 nm wide and a large strain field in W crystals; bubble coalescence generates larger bubbles as temperature increases up to 1500°C, while the surface partially recovers its original crystallographic structure. Finally, IsGISAXS simulations are presented to complete the description of the early stages of He bubbles formation, and their evolutions regarding the temperature and the crystallographic orientation of the sample.
[1] Y. Ueda et al., Fusion Engineering and Design 89 (2014) 7-8, 901-906.
[2] E. Bernard et al., J. Nucl. Mater. 484 (2017) 24-29
[3] E. Bernard et al., Nucl. Mater. En. 19 (2019) 403-409Speaker: Elodie Bernard (CEA Cadarache)
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11:45
Lunch
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RAMI&NT, H&CD sessionConvener: Didier van Houtte (CEA)
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Challenges for Engineering Plasma Facing Components for Fusion in Steady-State
The first wall and divertor of a fusion power plant (FPP) will experience heat, particle, and neutron fluences well beyond what has been seen for any existing plasma confinement experiments, and beyond what will be seen in ITER. Each plasma facing component (PFC) must satisfy numerous, often competing, requirements. Among the many challenges that must be addressed in designing these components are, neutron activation, particle sputtering and redeposition, erosion, helium implantation, swelling, fuel implantation, stresses produced by thermal gradients and coolant fluid pressure, fatigue from thermal cycling, creep from high temperature operation, etc. Most of these challenges will be present for regardless of which confinement concept is employed for the realization of fusion power. The need for a power plant to operate in long-pulse or steady-state with a high duty cycle will put requirements on these components that cannot be satisfied with current design solutions.
From an engineering standpoint, one issue that must be addressed is the ability to effectively remove heat from the component. The PFC can be conceptually divided into three aspects – the plasma facing (and interacting) surface, the coolant, and the intermediate thermal structure. Currently, many options are being considered for the plasma facing surface, and none are without problems and shortcomings. Tungsten is a commonly assumed candidate, but other materials, including liquid metals, could be explored. Several options are also being considered for the coolant. Water will be the primary coolant in ITER., but water has many limitations when considered for a fusion power plant, not least of which are: the possibility of tritium contamination, the high operating temperature of the wall, and the possibility of transmutation to Nitrogen 16. Helium, liquid metals and molten salt are being considered as alternatives, but further research is necessary before it will be clear that any will be an adequate coolant in a fusion power plant. Finally, the possible compositions and shapes of the intermediate thermal structure are too many to mention. It is clear that PFCs will need to be engineered components, probably several materials in alloy and/or unique shape.
Novel manufacturing, engineering, and simulation methods will be necessary to develop feasible technology for the plasma facing components in an FPP. This is emphasized in the recent US Community Plan: “Recognizing that it is unlikely that existing materials will provide adequate PFC system performance, it is imperative to initiate and sustain a program for the development of new, innovative solid materials that will form the basis of the solid first wall armor, solid divertor targets and the liquid metal PFC substrates through techniques such as advanced manufacturing, nano-engineered materials, material by design, virtual engineering” [1]. This presentation will consider innovations in topology optimization and advanced manufacturing that should enable the development of novel plasma facing components.
[1] Report of the 2019-2020 APS DPP Community Planning Process, A Community Plan for Fusion Energy and Discovery Plasma Sciences, https://drive.google.com/file/d/1w0TKL_Jn0tKUBgUc8RC1s5fIOViH5pRK/view, pg. 53, 2020.
Speaker: Arnold Lumsdaine (Oak Ridge National Laboratory) -
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Recent development of actively cooled PFCs for WEST, JT-60SA
In the current, and probably future, fusion devices, divertor is essential to achieve high performance plasma. In conjunction with the generic X-point configuration, its role is to absorb very intense heat flux located at the strike points, as well as keeping plasma impurities at a relatively low level.
For different reasons, such as melting temperature, erosion rate, He retention, tungsten is a material widely chose as main plasma facing material. The heat exhaust is then ensured by a heat sink generally made of CuCrZr, a copper alloy exhibiting acceptable mechanical properties, with water flowing through it. The fixation to the global divertor structure is done by steel parts. Such design with tungsten monoblock is used for ITER divertor, for which heat flux around 10MW/m², up to 20MW/m² on slow transients, may be expected. Its design and development of associated technological processes have lasted for more than ten years.
First part will show the design, manufacturing and qualification of this type of component for WEST lower divertor (Tore-Supra upgrade to a metallic machine), design to sustain pulse duration from 10s to 1000s. It has been equipped with 456 ITER grade components (called Plasma Facing Units – PFU), representing around 15% of the full ITER divertor area. Each PFU is made of 35 tungsten monoblocks assembled on a CuCrZr tube, via a copper interlayer ring, using hot isostatic pressure as assembly process. After the insertion of a twisted tape inside the tube, a stainless steel tube is welded by electron beam on both ends. Finally, the precise external shaping of the PFU is milled and grinded, including a 0.5mm / 1° bevel on the plasma interacting surface. This “fish-scale” shaping aims at protecting edges in case of misalignment, 0.5mm being also supposed for the bevel of ITER inner vertical targets. Finally, important surface modifications have been observed after a few campaigns in WEST, totalizing 4 hours and 25 minutes of plasma and 8.3GJ of deposited energy. Main observations are being reported, as a first step for future long pulse operations of fusion devices.
Second part will present current design studies regarding JT-60SA W divertor upgrade, expected to be operational in 2033. While the estimated maximal heat flux is also in the range 10MW/m² with possible excursions to 20MW/m² like in ITER and long pulse operations in the range of 100s, different environment (no D-T plasma) and different cooling conditions (from 90°C, 33bar, 10m/s to 40°C, 20bar, 7m/s) allow for a design different than ITER one. Recent development show that tungsten flat tiles assembled on a heat sink could be used, simplifying design and manufacturing processes. For heat sink material, several options are compared, including use of TZM, a molybdenum alloy capable of handling high temperature, or CuCrZr with a substantial increase of thermal capabilities using advanced design like hypervapotron combined with manufacturing process like additive fabrication. In addition to the power exhaust capacities, other specifications are taken into account, such as precise shaping, tolerance to the components misalignment and maintenance convenience.Speaker: Mehdi Firdaouss (CEA/IRFM) -
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Advance in heating and current drive by RF waves towards long-pulse operation on EAST
There are two lower hybrid current drive (LHCD) systems in EAST with the operating frequencies at 2.45 GHz (source power of 4 MW) and 4.6 GHz (source power of 6 MW), which is the main electron heating and current drive (H&CD) tool. Hot spots issue and anomalous loss of LHCD efficiency at high density are the big challenges in long-pulse and high betap operation with high LH power. By upgrading the guard limiters, the power handling capability of LH systems are improved and the 4.6 GHz LH system can operate with power > 2.5 MW routinely. With lithium wall conditioning and favourable magnetic field, the interactions of the LH waves with the scrape off layer (SOL) plasmas are mitigated, indicated by the frequency spectral measurements. As a result, good LHCD efficiency ~ 0.9 *10^19 A/W/m2 is obtained even when density increased to 5.4 * 10^19 /m3 in H-mode plasmas (the Greenwald density fraction ne / nG ~ 0.75). In order to decrease the interactions of LH antenna with SOL plasmas, a new passive-active-multijunction (PAM) launcher has been developed for the 2.45 GHz system, and good coupling with reflection coefficient ~ 3% has been demonstrated with the plasma-antenna distance up to 11 cm. High electron temperature Te >12 keV plasmas measured by Thomson scattering is produced by injecting 1.4 MW electron cyclotron (EC) waves into the centre of plasmas sustained by 2.3 MW LH waves. With these improvement, significant advance has been achieved with RF waves, including long-pulse H-mode operation up to 312 s with 1.8 MW 4.6 GHz LH power, 1.6 MW EC power and 1.2 MW ion cyclotron (IC) power, long-pulse I-mode operation up to 1056 s sustained by moderate 4.6 GHz power ~ 1.05 MW and EC power ~ 0.55 MW. In order to enhance the H&CD capability, the 2.45 GHz system will be replaced by a new 4.6 GHz / 4 MW system with a PAM launcher, which is under construction.
Speaker: Miaohui LI (Institute of Plasma Physics,Chinese Academy of Sciences (ASIPP)) -
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Optimization of the beam heating and non-inductive current in NBI plasma for long-pulse operation of EAST
In fusion devices, the first wall will be hit by lost fast ions and release large amounts of impurities into plasmas, which will affect the operation of experiments. Thus, the beam ions needs to be optimized for the safety and long-pulse operation in neutral beam injection (NBI) plasma.
In our research, the beam heating and loss, neutral beam current drive (NBCD) have been numerically investigated by using NUBEAM, ONETWO and TRANSP. The simulation results show that the heating efficiency is improved by optimizing the tangential radius and injected direction. For co-NBI, the heating efficiency is enhanced with the increasing of tangential radius. Moreover, the prompt loss is obviously reduced when the beam injected direction changes from counter-current to co-current. For the plasma with $\langle n_e \rangle$ ~ 3.510$^{19}$ $m^{-3}$ ( $f_{GW}$ ~ 0.45), the heating efficiency is nearly doubled due to this direction optimization. There is good agreement between the experimental neutron amplitude and the calculated neutrons by using TRANSP/NUBEAM. For NBCD, its efficiency increases with the beam power and tangential radius. By changing the beam injected direction from counter- to co-, the efficiency raises nearly two orders of magnitude. Thus, the efficiency is shown to be about 0.02 for a well optimized system on EAST.
Furthermore, by changing the gapout, the magnetic configuration has been optimized for higher heating efficiency in different collisional plasmas. The simulation results show that the expansion of gapout helps to improve the beam heating efficiency due to the power deposition moving inward in high collisional plasma. Meanwhile, the fraction of trapped particles is decreased.Then the prompt loss and ripple loss of fast ions are reduced. In low collisional plasmas, the gapout needs to decrease to reduce the shinethrough loss and improve the beam heating. Thus, the gapout needs to be optimized based on the collision rate. Moreover, the gapout affects the current profile. The simulation results illustrate the profiles of NB driven current and bootstrap current move inward with the increasing of the gapout. In EAST discharges 101726 and 101731, the only difference in the magnetic equilibrium is the gapout varing from 8.4cm in #101726 to 6.6cm in #101731. After injected similar deuterium beams, the neutron amplitude and the internal inductance li are both higher in #101726 (gapout=8.4cm) than that in #101731 (gapout=6.6cm) . In the recent long-pulse NBI discharge 104243 ( $\langle n_e \rangle$ ~ 4.010$^{19}$ $m^{-3}$ ) with about 60s, the gapout is about 9cm which is higher than the normal gapout (~ 6-7cm) on EAST. This suggests the bigger gapout is benefit for the long-pulse operation in high collisional plasma.
In the end, since the beam also has influence on the bootstrap current by affecting temperature profiles. Therefore, we optimized the bootstrap current by varing beam energy, density and composition of the plasma for high $β_p$ and fully non-inductive discharges in long-pulse NBI plasma on EAST.
Acknowledgments:This work was supported by the National Key R&D Program of China under Grant (No.2017YFE0300404,2019YFE03020004) and National Natural Science Foundation of China (No. 12175272,11875290 ).Speaker: Jinfang Wang (Institute of Plasma Physics Chinese Academy of Sciences) -
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Potential of lower hybrid fast wave as an efficient current drive method in future fusion reactor
An efficient current drive method should be developed for future tokamak reactors since bootstrap current is not able to soley fufill the plasma current required for enough confinement for burning plasma and current profile should be carefully controlled for stable operation. Lower Hybrid Current Drive (LHCD) has been typically considered as an current drive method of tokamak due to its supreme efficency. However, its high efficiency has not been shown in the reactor relevant high density plasma. Currently, helicon wave current drive and high field side launch LHCD have been suggested and researched in the several tokamaks such as KSTAR, DIII-D to overcome the difficulty. However, its theoretical good efficiency of helicon wave has not been shown experimentally yet, and it is not obvious the high field side LHCD is capable of efficient current drive still for high density plasma. Meanwhile, the fast wave in lower hybrid resonance frequency, so called, Lower Hybrid Fast Wave (LHFW) has a good penetration characteristic in the high density and its current drive efficiency is predicted to be as efficient as lower hybrid slow wave used in LHCD. Recently, it is shown that the current drive is possible if the density window for propagation is suitably open on VEST deice. In the study, the propagation and current drive characteristics of LHFW in ITER or DEMO are investigated through ray tracing simulation. The coupling and power transmission of LHFW is estimated with a fast wave antenna using waveguide under conceptual design for reactor grade tokamak. Based on those results, the potential of the LHFW as an altenative current drive method for the future fusion reactor is discussed and summarized.
Speaker: Sun Ho Kim (Korea Atomic Energy Research Institute, Daejeon, Korea) -
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Research on high current and low inductance laminated transmission busbar in high-power long-pulse steady-state operations
Abstract—The operation goal of future fusion reactors, such as China Fusion Engineering Experimental Reactor (CFETR), International Thermonuclear Experimental Reactor (ITER) is to realize high-power long-pulse steady-state plasma. The conventional DC busbar occupies too much space, and the large inductance leads to voltage drop, increase EMI (Electromagnetic Interference) and no on-line real-time diagnosis and detection of its state is not suitable to the long-pulse steady-state operation.
This paper presents an improved high current transmission busbar available for long-pulse steady-state operation. The fully insulated laminated transmission busbar with low inductance, low impedance, high power density and high current will be studied and analyzed. Three formulas are established to optimize the low inductance and insulation level. With same current and voltage, the new high current laminated transmission busbar can reduce 70% of the installation space, 50% of the stray inductance. Moreover, the contact resistance the transmission loss can be reduced by 20%. It can correspondingly reduce the current rise time and suppress the long-pulse current ripple.
To make the on-line real-time diagnosis during long-pulse operation, firstly, the influence of the oxide layer on the contact surface and the electrical contact resistance is analyzed. Secondly, the on-line real-time diagnosis and detection of the insulation performance of the first and second conductors of the laminated transmission busbar are studied, in which the electrical contact is analyzed for fatigue and aging test. The aging test could provide the basic information to determine for threshold of transmission busbar lifecycle. Finally, the experimental test and analysis are carried out, and the calculation results are compared with the theoretical analysis to verify the effectiveness of the method and prove the main conclusions of this paper.Keywords - laminated transmission busbar; low impedance; low inductance; stray inductance; high insulation level; high power density; long-pulse; contact resistance.
Speaker: Zhengyi Huang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences) -
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Preliminary design on power supply system of superconducting test device based on long pulse signal
To meet the requirements of continuous improvement of long pulse parameters, it is necessary to build a high-power converter power supply system, and optimize and analyze the power supply system based on the previous EAST (Experimental advanced superconducting Tokamak) and ITER (International Thermonuclear Experimental Reactor) fusion devices. This paper takes the power supply system of CRAFT (Comprehensive Research Facility for Fusion Technology) fusion device as the research object, and studies the performance optimization of various equipment from the perspective of adapting to long pulse. The converter power supply system is mainly composed of high-voltage AC switchgear, converter unit, and isolation network of combined converter to realize various operation modes, converter power control system and protection system. In this paper, the main converter unit is designed, including the calculation of three winding phase-shifting rectifier transformer, the structure analysis of thyristor converter, the parameter research of DC reactor, etc. The results have the characteristics of low impedance, high efficiency and stability, and can ensure the stable output of long pulse. The research in this paper provides an urgent research, development and testing platform under extreme conditions for the future research, development and testing of high-power fusion, special power supply and DC switchgear.
Speaker: Ya Huang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)
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15:40
Coffee
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CICLOP sessionConvener: Tomohiro Morisaki (National Institute for Fusion Science)
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Long plasma duration operation analyses with an international multi-machines (tokamaks and stellarators) database
Combined high fusion performance and Long Pulse Operation (LPO) is one of the key integration challenges for fusion energy development in magnetic devices. Solving this problem requires a comprehensive vision of the physical and engineering aspects to simultaneously increase the duration and performance of fusion. Significant progress has been made in tokamaks and stellarators, including very recent advances in duration and/or performance. These progresses are reviewed by analyzing the experimental data (109 pulses / 3200 entries) provided by 10 tokamaks (ASDEX Upgrade, DIII-D, EAST, JET, JT60-U, KSTAR, TCV, TFTR, Tore Supra, WEST) and two stellarators (LHD and W7-X) expanding the pioneering work of M. Kikuchi [Frontiers in Fusion Research, Springer]. Data have been gathered and coordination has been provided by the recently created IEA-IAEA international CICLOP group (Coordination on International Challenges on Long duration OPeration).
Using the international multi-machines database, the latest achievements will be reviewed in terms of input energy (for example, 1730 MJ in L-mode, 425 MJ in H-mode), duration (1056s in L-mode, 101s in H-mode), input power and sustained performance. Progress has been made in maintaining the LPO in tokamaks and stellarators with superconducting coils, actively cooled components and/or with metal walls. The graph of the dependence of the triple fusion products on the duration shows a sharp decrease by at least two orders of magnitude with an increase in the plasma duration from less than one second to 100s. Indeed, LPO is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3keV for discharges above 100s. Difficulties in extending the duration may arise from coupling high heating powers over long duration and the evolving plasma-wall interaction towards an instable operational domain. Possible causes limiting the duration will be reported and analyzed as critical issues to be addressed prior ITER operation and DEMO design.Speaker: Dr xavier Litaudon (EUROfusion)
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Discussion session and Closing
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