Second Technical Meeting on Long-Pulse Operation of Fusion Devices

14 Oct 2024, 08:30 → 18 Oct 2024, 17:15 Europe/Vienna

Board Room D (IAEA Headquarters)

Xavier LITAUDON (CEA)

KEY DEADLINES

3  May 2024 Deadline for submission of abstracts through IAEA-INDICO for regular contributions

31 July 2024 Deadline for submission of Participation Form (Form A),  and Grant Application Form (Form C) (if applicable) through the official channels

14 June 2024 Notification of acceptance of abstracts and of assigned awards

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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.

Monday, 14 October

10:00 → 12:00 Registration

12:00 → 14:00 Lunch Break

14:00 → 14:30 Welcome and Introduction

Conveners: Mr Matteo Barbarino (International Atomic Energy Agency), Xavier LITAUDON (CEA)

14:30 → 15:40 LPO session

14:30 The new ITER Baseline research plan and long-pulse/steady-state operations in ITER

The ITER research plan has been recently re-elaborated to lay out a robust path to achieve the Project’s fusion production goals and technical objectives. This includes changes to device and ancillaries configuration to minimize previously identified risks such as: changing plasma facing material of the first wall from beryllium (Be) to tungsten (W), modifying the heating and current drive systems mix as well as the phased introduction of water cooled components for the first wall. The initial phase of the operation, so-called Start of Research Operation (SRO) phase, will utilize an inertially cooled wall to minimize operational risks, address access to deuterium H-mode operations with pure electron heating leading to low neutron production and demonstrate the full technical capabilities of the tokamak at full field and current, including effective disruption mitigation. The first Deuterium-Tritium phase (DT-1) focuses on the achievement of the Project’s Q=10, 500MW with burn duration of, at least, 300 s within ~ 1% of the present neutron fluence objective of ITER. This will provide a scientific basis on burning plasmas as well as key information on the technical performance of the ITER systems, as necessary inputs to licencing for DT-2. The second Deuterium-Tritium phase (DT-2) will progressively demonstrate the Q≥5 long-pulse and steady-state operation with 1000s and 3000s of burn duration respectively, as strong candidates for plasma scenarios in future fusion power plants. Tritium breeding technologies will also be addressed by implementing the ITER test blanket module (TBM) program throughout the DT-1 and DT-2 phases. This contribution will describe the new ITER baseline research plan with an emphasis on the development of ITER long-pulse and steady-state operations, as well as open R&D issues and required activities.

Speaker: Sun Hee KIM (ITER Organization)

15:05 Recent Experimental Results on EAST in Support of ITER and CFETR SSO

Significant progress has been achieved on EAST in the development of long-pulse steady-state advanced plasmas in support of future fusion reactors (e.g ITER and CFETR) since last IAEA technical meeting (TM) on long-pulse operation (LPO). A new record of reproducible 403s H-mode plasmas has been achieved with full metal wall with EAST improved flexibilities and capabilities. EAST scientific objectives focus on physics understanding for three scenario regimes attractive for ITER baseline, hybrid and steady-state operations. EAST experimental results have demonstrated enhanced heating and current driven efficiency by using high launched frequency 4.6GHz LHW system and ICRF antennas with lower parallel wave number k|| combined with on-axis/off-axis ECRH, MHD and transport behaviour at zero loop voltage, long-pulse fuelling and recycling control by pellet injection/SMBI and real-time wall-conditioning, improved heat exhaust with new type W/Cu divertor to avoid local hot-spots, good energy confinement exhibited in high poloidal beta scenario at high density, robust plasma shape control by upgrading magnetic measurements and optimizing plasma ramp-up/down phase, etc.
Recently, to support ITER’s proposed new research plan, a dedicated set of joint ITER-EAST experiments have been performed, and the related key technical and scientific challenges have been addressed. For wall conditions with boron coating, to improve the uniformity and quality of the boron film, optimization and characterization of boronisation using ICWC and GDC has been investigated. For plasma initiation, ECW assisted start-up has been carried out and robust breakdown and plasma initiation at low toroidal electric fields has been demonstrated with optimized magnetic field configurations in a large range of prefill gas pressure. In boron wall, we demonstrated a stationary 100s H-mode plasma, and also achieved high ion-temperature plasma up to 100 million ºC accompanied with fishbones using impurity (Ar) seeding. A set of dedicated experiments has been carried out to assess the impact of W as the first wall material in H-modes. A long-pulse (>50s) no-ELM H-mode discharges was achieved with feedback-controlled divertor detachment via nitrogen seeding. ELM control by various methods (RMP, impurity powder, etc) has been further studied in different scenarios. The impact of different wall conditions (lithium-coating, boron-coating, non-coating) on confinement, H&CD efficiency, particle control, etc. will be illustrated in detail, which can offer unique contributions in support of the ITER new research plan.

Speaker: Juan Huang (CnIPPCAS)

15:40 → 16:10 Coffee Break

16:10 → 17:20 LPO session

16:10 Long pulse operation in a W environment : feedback from WEST

Long pulse operation in a W environment : feedback from WEST
P. Manas1, P. Maget1, R. Dumont, J. Dominski2, T. Fonghetti1, J. Morales1, A. Ekedahl1, N. Fedorczak1, J. Gaspar3, E. Tsitrone1, and the WEST Team*
1 CEA, IRFM, F-13108 Saint-Paul-lez-Durance, France
2 Princeton Plasma Physics Laboratory, 100 Stellarator rd, Princeton 08540 New Jersey,USA
3 Aix-Marseille Université, CNRS, IUSTI, Marseille, France
* see http://west.cea.fr/WESTteam

Record pulses in terms of injected energy and pulse duration at the superconducting WEST tokamak have been recently achieved. These have been obtained with an ITER-grade actively cooled divertor and with lower hybrid heating and current drive for up to 6 min and with 1.15 GJ of injected energy [1]. These long plasmas evidence the importance of material ageing, tungsten contamination and boronisation impact. A High Fluence campaign in attached divertor regime (cumulating 3 h of plasma, and reaching a divertor particle fluence equivalent to 1 ITER shot) demonstrated the formation of tungsten deposits, and their release under the form of UFOs impacting plasma operation [2, 3]. This process equilibrates after about 2h of plasma, with a rate of UFO generation saturating at about 6 UFO per minute of plasma. Actuators for coping with UFOs are essential, as their occurrence becomes more probable with extended pulse duration. Some of the few remote vessel elements that are not actively cooled were shown to outgas H species when approaching the GJ range, and to be conditioned progressively as long pulses were repeated, thus evidencing the need for an exhaustive cooling of the in-vessel components. Detached scenarios with nitrogen seeding (in particular the X-point radiator regime [4] where feedback control has been demonstrated for more than 10 s and also obtained in double null configurations) are currently developed to improve on two important long pulse aspects by (i) limiting the erosion of the divertor possibly limiting the level of UFOs observations (ii) improving the plasma performance in terms of confinement time and ion temperature via turbulence stabilisation. Additional beneficial effects were also observed on the tungsten core contamination, which robustly yields ~50% fraction of core radiated power in standard scenario and can be reduced down to ~40%. While the tungsten central peaking remains modest in these RF heated plasmas [5], increasing nitrogen content also reduces the central neoclassical W peaking [6]. New perspectives for WEST contributions to ITER include the extension of the low divertor temperature regime to Long Pulse Operation (over one minute), the characterization of W limiter operation (coordinated by ITPA), as well as further addressing high-injected energy challenges.
References
[1] P. Maget et al, EPS 2024 [2] E. Tsitrone et al, PSI 2024
[3] J. Gaspar et al, PSI 2024 [4] N. Fedorczak et al, PSI 2024
[5] X. Yang et al, NF 2020 [6] J. Dominski et al, APS 2023

Speaker: Pierre Manas (CEA, Cadarache)

16:45 Achievement of 102-second High-performance Long-pulse Discharge with Lower W-shaped Tungsten Divertor in KSTAR

The mission of the KSTAR device is to sustain high-performance plasma for 300 seconds[1]. In the 2023 experimental campaign, KSTAR maintained H-mode plasma for 102 seconds. The operating conditions of this discharge were $I_P$=400 kA, $B_T$=1.95 T, $P_{NBI}$=3.9 MW, and $P_{EC}$=1.1 MW. The plasma characteristics of this plasma were $V_{loop}$~70 mV, $β_P$~2.5, $β_N$~2.1, $T_{e,core}$>6.0 keV, $T_{i,core}$~2.5 keV, and $n ̅_{e,core}$~3.0x$10^{19}$ $m^{-3}$. The achievement of high-performance 102-second plasma was based on the following factors. Firstly, to effectively control heat flux to the divertor and protect the inner walls of the device, KSTAR installed a W-shaped tungsten divertor capable of active water-cooling. The temperature variation measured by thermocouples on the tungsten divertor during the 102-second plasma was less than 15 °C, approximately 1/25th compared to previous long-pulse discharges. However, other PFCs excluding the tungsten divertor showed temperature variations of up to 100 °C. Secondly, an algorithm was successfully developed to real-time correction of the linear signal drift of magnetic diagnostics in PCS and applied to long-pulse plasma experiments. During the 102-second long-pulse plasma, key plasma shape variables were effectively controlled within a maximum error of 2 cm. Thirdly, to maintain a relatively low line-averaged electron density for lower loop voltage, the plasma shape scenario was updated. Changes in H-mode characteristics in the W-shaped tungsten divertor environment were monitored due to changes in the plasma-facing material and divertor geometry, which influenced the plasma shape and the state of the SOL region. Consequently, H-mode characteristics in the partially covered tungsten environment appeared to be not significantly different from those in the previously fully covered carbon environment. However, in the H-mode plasma at $I_P$=400-500 kA without additional gas injection, the line-averaged electron density was maintained at a higher value of ~4.0 x$10^{19}$ $m^{-3}$ compared to previous discharges. To enable long-pulse plasma operation, it was necessary to reduce plasma density, which was partially achieved by adjusting the plasma shape. Fourthly, the phenomenon of gradual plasma performance degradation over time was significantly alleviated. The performance degradation that typically occurred around 20 seconds[2] was observed to be minimal up to approximately 70 seconds, where plasma performance remained almost constant. This is likely to have been influenced by changes in divertor geometry affecting the state of the SOL region and the appropriate scenario of gas injection. A plasma discharge lasting for 300 seconds should be pursued in conjunction with the development of a fully non-inductive plasma scenario. This work is expected to contribute to the development of scenarios aimed at maintaining consistent plasma performance over time and monitoring the temperature behavior of PFCs during long-pulse plasma operation.

This work was supported by the R&D Program of “KSTAR Experimental Collaboration and Fusion Plasma Research (EN2401-15)” through the Korea Institute of Fusion Energy (KFE) funded by Government funds.

References
[1] G.S. Lee, et al., Nuclear Fusion, 40, 575 (2012)
[2] H.-S. Kim and Y.M. Jeon et al., Nuclear Fusion, 64, 016033 (2024)

Speaker: Dr Hyun-Seok KIM (Korea Institute of Fusion Energy)

Tuesday, 15 October

09:00 → 10:35 LPO session

09:00 Development of long pulse scenarios at the stellarator Wendelstein 7-X

Stellarators offer a relatively straightforward route to steady state fusion power operation. Future reactors, such as DEMO, are expected to operate quasi-continuously with divertor heat fluxes as low as 5 MW/m$^2$ [1]. This requires that most of the power leaving the plasma has to be radiated away by the plasma impurities in the so-called detached regime, in which the divertor target plates are protected from the hotter plasma by a dense layer of cold (T$_e$ < 5 eV) plasma. One of the objectives of the Wendelstein 7-X research is to achieve and optimise the long pulse scenarios. The recently installed high heat flux divertor, consisting of water-cooled CFC target elements, can handle power loads of up to 10 MW/m$^2$ in continuous operation, which for W7-X plasmas means up to 30 minutes.
During the W7-X operational phase OP1.2 (2017-2018), discharge lengths of 100 s with attached [2] and 30 s with detached plasma [3] were achieved. In the latter case, the detachment was successfully stabilised by keeping the plasma density at a relatively high level (~1.2$\times$10$^{20}$ m$^{-2}$) necessary to set the radiation fraction to approximately 90% of the input electron cyclotron resonance heating (ECRH) power. The scenario was developed using feedback-controlled gas fuelling with the line-integrated electron density and total plasma radiation as control parameters [4]. The experiments showed that a robust detachment scenario allows to reduce the peak heat flux by almost one order of magnitude, and no significant increase in impurity concentration was observed.
In the recent campaign OP2.1, we significantly extended these scenarios reaching in attached conditions plasma durations of up to 8 min in the attached state (ECRH input power of 3.5 MW; line-integrated electron density was set to approximately 5$\times$10$^{19}$ m$^{-2}$ and kept constant throughout the discharge with a density feedback system; the diamagnetic energy was kept at a level of about 430 kJ, and the ion and electron temperatures were 1.5 keV and 2 keV, respectively). Such a scenario did not lead to any overloading of the surface of the components facing the plasma.
In detached case, a more sophisticated scenario had to be developed which resulted in up to 2 min. discharge lengths with almost completely detached divertor (surface temperatures of 150-160°C). ECRH input power of 5 MW was applied, both intrinsic (mostly carbon) and seeded (neon) impurities were used to increase the plasma radiation fraction up to ~80%). The line-integrated electron density was set at 1.3$\times$10$^{20}$ m$^{-2}$, resulting in both ion and electron temperatures at the level of 1.5 keV and a diamagnetic energy of about 600 kJ. No increase of impurity concentration was observed with line-averaged Z$_{eff}$ values below 1.5 and stable line emission of e.g. carbon and oxygen in the X-ray spectra of the Pulse-Height Analysis diagnostic.
[1] N. Asakura et al., Nuclear Fusion 57, 126050 (2017).
[2] T. Klinger et al., Nuclear Fusion 59, 112004 (2019).
[3] M. Jakubowski et al., Nuclear Fusion 61, 106003 (2021).
[4] M. Krychowiak et al., Nuclear Materials and Energy 34 101363 (2023).

Speaker: Maciej Krychowiak (Max-Planck-Institute for Plasma Physics)

09:35 High density and high neutral particle pressure in the divertor for steady state operation in LHD

This study shows the experimental results on the density limit and ultra-high neutral pressure in the sub-divertor volume at LHD for high density steady state operation, and good particle exhaust.

A new density scaling for tokamaks was reported in [1] based on edge turbulent transport. It shows a strong dependence on heating power and is similar to the Sudo density limit in the helical device, LHD [2]. In LHD, it is possible to observe a large number of density collapse events with the measurement of microscopic turbulence such as phase contrast imaging (PCI) and Doppler back-scattering (DBS). Preliminary results show that the density fluctuation level increases at the density collapse, and this study focuses on the relation between the density limit and edge turbulence.

Ultra-high neutral pressures were recently observed in the helical divertor in LHD [3]. The record value was 1.4 Pa, which is in the order of magnitude of the values known from poloidal divertors of tokamaks (e.g. 5 Pa in ASDEX Upgrade or 20 Pa in ITER). This result has been observed only at a specific magnetic field configuration of Rax = 3.55 m with an inward directed magnetic axis position. The high divertor pressure is probably caused by volume recombination. In general, volume recombination can be easily achieved in tokamak divertors, but this has not been the case for stellarator/heliotron divertors. Therefore, these results have not been seen in previous stellarator/heliotron divertors, and the process will be discussed in this study.

[1] M. Giacomin et al., Phys. Rev. Lett. 128 (2022) 185003.
[2] S. Sudo et al., Nucl. Fusion 30 (1990) 11.
[3] U. Wenzel et al., Nucl. Fusion 64 (2024) 034002.

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

10:00 High density as an avenue towards high confinement quality and core-edge integration in advanced tokamaks

Recent high poloidal beta ($\beta_P$) scenario experiments on EAST and DIII-D have made coordinated breakthroughs for high confinement quality at high density near the Greenwald limit. Experiments on DIII-D have achieved $f_{Gr}$ ($=$line-averaged density$/$Greenwald density) above 1 simultaneously with $H_{98y2}$ around 1.5, as required in compact steady-state fusion pilot plant designs but never before verified in experiments. Compatibilities of high confinement core with small ELMs and fully detached divertor have been demonstrated separately in DIII-D high $\beta_P$ experiments with $f_{Gr}\sim1.0$. Experiments on EAST have nearly doubled the ion temperature at $f_{Gr}\sim0.9$, confirming the predict-first results of simulations. Density gradient amplification of turbulence suppression at high $\beta_P$ is the underlying physics that can explain, in both devices, the achievements of improved confinement at high density.

EAST long pulse H-mode experiments have reached a world record duration of 400 seconds, but with $T_i\ll T_e$. Transport modeling indicated that the ions are limited by ITG modes. Modeling also suggested potential solutions, including reduced magnetic shear, and enhanced density gradients. Following this guidance, various approaches were pursued on EAST. The experiments directly show that a strong enhancement of Ti happens with a single short pulse (100 ms) of impurity injection at $f_{Gr}\sim0.9$. But it can only happen in the appropriate conditions, i.e. a combination of low magnetic shear and high density gradient, as predicted by the earlier modeling.

On DIII-D, a synergy between increased $H_{98y2}$ and increased $f_{Gr}$ is observed in low-$l_i$ plasmas with strong gas puffing, due to the build-up of an internal transport barrier (ITB) at large radius in the temperature and density channels. Sustained $1.0Nature (2024), https://doi.org/10.1038/s41586-024-07313-3]. The experimental approach for high $f_{Gr}$ is to elevate the core density by developing a strong ITB, while keeping the pedestal density below the Greenwald limit. Transport simulations show lower turbulent energy transport at higher density gradient due to stronger $\alpha$-stabilization effect. Simulations also reveal that the favorable trend of low transport at high density is only expected when increasing the density gradient at high local safety factor ($q$) and high $\beta$, thus at high $\beta_P$ to ensure strong turbulence stabilization. Sustained small ELMs and reduced divertor heat load are observed simultaneously in the $f_{Gr}>1.0$ and $H_{98y2}\sim1.5$ phase. Increased separatrix density is believed to be a part of the physics that leads to realization of the small-ELM regime. Excellent compatibility of actively controlled full divertor detachment with high density and confinement ($f_{Gr}\sim0.9$, $H_{98y2}\sim1.5$) has been demonstrated in separate DIII-D high P experiments [L. Wang, Nat. Commun. 12 (2021) 1365]. Despite decreased pedestal pressure when entering detachment, a stronger ITB, facilitated through self-organization, leads to improved confinement, in contrast to the confinement degradation with divertor detachment commonly observed in standard H-modes. These results demonstrate the possibility of integrating excellent plasma confinement at high density with an efficient divertor solution, an essential step towards steady-state FPPs. Supported by the US DOE under DE-FC02-04ER54698 and DE-SC0010685.

Speaker: Siye Ding (General Atomics)

10:35 → 11:05 Coffee Break

11:05 → 11:30 LPO session

11:05 Long pulse operation with the JET ITER-Like Wall

In the last experimental campaign of the JET tokamak in December 2023, long discharge operation (>30s) in deuterium plasmas was developed to assess the sustainment of the plasma performance over many resistive time scales and to address plasma-wall interaction physics in a full metallic environment with the ITER-like wall (ILW), with a W divertor and a Be first wall [1]. Two types of long duration discharges were successfully developed for this purpose: (i) a 30s ELMy H-mode with combined 12-14MW neutral beam heating (NBI) and 2MW of ion-cyclotron resonance heating (ICRH) and (ii) a 60s long pulse with 4-5MW of NBI and 2MW of ICRH. Both operational scenarios are based on previously developed hybrid-like plasmas at JET [2] with IP=1.4MA and B0=1.9T (q95~4), the latter being compatible with central hydrogen minority ICRH at 29MHz. The 30s pulses had an averaged stored energy of WP=2.5MJ (betaN~2), with ne0=3.5e19/m3 and core temperatures of Te0~Ti0=4-5keV. The pulses were stationary from the radiation point of view with no sign of core impurity accumulation and featured regular type-I ELMs throughout the discharge with frequency fELM~100Hz. The 60s pulses required significant technical adjustments in many subsystems, ranging from plasma shape control, machine protection and diagnostic systems settings [3]. A stationary 60s heated discharge was achieved for the first time in JET-ILW (the last of such pulses been done in the 90’s with a C-wall and different divertor structure), with ne0=4.0e19/m3 and Te0~Ti0=3.5keV (Wp=1.5MJ). This pulse achieved the maximum energy ever injected in a single pulse in JET, Ein=390MJ, and challenged the operational domain of the (passively cooled) divertor limits, reaching Ediv=315MJ. An overview of the main properties of these discharges in terms of stationarity, plasma-wall interaction and overall performance in the context of the international CICLOP database [4] will be presented. A more detailed analysis of the different topics will be given in dedicated papers [5].

[1] G. F. Matthews et al 2011 Phys. Scr. 2011 014001
[2] J. Hobirk et al 2023 Nucl. Fusion 63 112001
[3] D. King et al, “Technical and Engineering challenges for long pulses on JET ITER
Like Wall”, this meeting
[4] X. Litaudon et al 2024 Nucl. Fusion 64 015001
[5] S. Bresinzek et al, ..., this meeting

Speaker: Dr Ernesto Lerche (Laboratory for Plasma Physics, ERM/KMS and UKAEA-CCFE Culham Science Centre)

11:30 → 12:30 PWI session

11:30 Global particle balance investigation using hot wall operation in all-metal plasma facing device, QUEST

QUEST is a medium-sized spherical tokamak which has parameters of R~0.64m, a~0.4m, BT<0.25T. The plasma facing walls (PFWs) are made of metals such as tungsten, atmospheric sprayed tungsten (APS-W), stainless steel type 316 (SS 316L) and 316L (SS316). A RF heating source is available for long pulse operation (LPO). A hot wall has been installed to regulate PFWs temperature. Taking advantage of the equipment, several times of LPO have been performed up to 6 hours.
In this study, we try to investigate the quantitative wall pumping and/or exhaust capability for various PFWs used in QUEST including plasma-induced deposition layer (DL). Tungsten is the most promising materials for ITER and mono-block tungsten has been used for limiters on QUEST. Its characteristics of hydrogen pumping and/or exhaust capability has been strenuously investigated in various fusion-oriented devices. APS-W is a coating of the hot wall facing to plasmas. SS316L is used for electrode of coaxial helicity injection (CHI) locating on the bottom side. SS316 panels has been installed since 2018 as a protection of the vacuum vessel in the high filed side from plasma bombardments. The SS 316 panels cold be cooled by water. According to the quantitative investigation, it found that the SS316 panels played an essential role in global particle balance during LPOs. Tungsten and APS-W have quite low solubility of hydrogen and their wall pumping capability is significantly limited. But a part of APS-W was covered with plasma induced deposition layer (DL) composed of carbon, tungsten, and stainless-steel materials and has a larger capability of wall pumping capability than original APS-W. The impact of the DL is also clarified quantitatively using the fast ejecting system for targeted sample (FESTA) measurements. The hydrogen desorption from the DL is approximately 10 times larger than SS316L and it means that the DL could play a significant role in global particle balance. Actually, the regulation of the hot wall temperature gave rise to extension of plasma duration from 40 min to more than 3 hours.

Speaker: kazuaki Hanada (Advanced Fusion Research Center, Research Institute for Applied Mechanics, Kyushu University)

12:05 Hydrogen retention and outgassing analysis with examples from JET-ILW, including long-pulse discharges

Plasma-wall interaction (PWI) plays a crucial role in plasma control in fusion devices. Apart from impurity generation and power exhaust issues, fuel recycling at plasma-facing components (PFCs) introduces additional challenges by affecting plasma density control and fueling efficiency. Hydrogen transport, retention and release from PFCs is influenced by the incoming ion and neutral particle fluxes, surface temperature and material properties. Thus fuel wall pumping and release vary both in location, as different PFC elements are exposed to different plasma conditions, and in time into the discharge, due to changes in fueling, surface temperature and, potentially, material microstructure.
Over 40 years of operation, JET accumulated a vast database on PWI, including deuterium-tritium (DT) campaigns and operation with the ITER-like wall (ILW), beryllium first wall and tungsten divertor, providing invaluable insights into fuel retention and recycling processes. In November and December 2023, almost at the very end of JET operations, several 30s and 60s long stationary heated discharges (compared to typical ≤20s long discharges) were performed in JET-ILW [1, 2]. These specially developed extended discharges offer the opportunity for investigation of the pathway to plasma-wall equilibrium in terms of fuel retention and recycling. This contribution will briefly summarize the main key points and outcomes in view of fuel retention from PWI experiments in JET and present a comparison of conventional and long-pulse discharges in JET-ILW by in-vessel gas balance and post-discharge outgassing analysis. In particular, preliminary intra-shot gas balance analysis indicates faster wall saturation in high power (14-16 MW) 30s long pulses compared to lower power (6-7 MW) 60s pulses, in which wall saturation effects take onset only after 10-20s.
[1] D. King et al., Technical and Engineering challenges for long pulses on JET-ILW, this meeting
[2] E. Lerche et al., Long pulse operation with the JET ITER-Like Wall, this meeting
* For the list of JET contributors see C.F. Maggi et al., Overview of T and D-T results in JET with ITER-like Wall, accepted for publication in Nuclear Fusion (DOI: 10.1088/1741-4326/ad3e16)
** For the EUROfusion Tokamak Exploitation Team see E. Joffrin et al., Progress on an exhaust solution for a reactor using EUROfusion multi-machines capabilities, accepted for publication in Nuclear Fusion (DOI: 10.1088/1741-4326/ad2be4)

Speaker: Dr Dmitry Matveev (Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – IEK-4, Jülich, Germany)

12:30 → 14:00 Lunch Break

14:00 → 15:15 PWI session

14:00 Effect of boron coating on long pulse high confinement plasma in EAST with full metal wall

Boron, as a low-Z material, is widely employed for wall conditioning to enhance plasma performance in fusion devices. Boron coatings including pre-discharge coating by using carborane (C2B10H12) as the working material assisted by ion cyclotron wall conditioning (ICWC) and real time coating have been successfully performed in EAST machine with full metal first wall [1].
After pre-discharge boronization, it was found the thickness of B film was about tens to one hundred nm and the surface of the sample was granular. The main composition of B film was about 50% B, 30% C and other elements including O, N, and W analyzed by XPS. The impurity radiation including oxygen and heavy impurities such as W, Fe, Cu and Zeff decreased significantly, which results in the slightly increased plasma stored energy. The lifetime of boronization was about 1700s in EAST. However, the H release was very serious during the initial plasma discharges after boronization due to H co-deposition during boronizaiton. To avoid introducing H isotopes, pure B powder with an average size of 70 µm was injected into plasma for real time boron coating. The reduction of the low-Z and high-Z impurities were observed[2], and the W impurity content could be decreased to 10-5 as the boron powder continuously injecting. Furthermore, it was found that the fuel particle recycling decreased with an increase in the amount of B powder injected. The fuel recycling decreased by up to 80%, and each B atom exhibited a trapping capacity of 0.3 D particles during B powder injection at a typical flow rate of 20 mg/s by particle balance analysis. The possible mechanism for D retention is the formation of B-C-O-D compounds and co-deposition between B and D particles during discharges [3] .
By performing these boron coatings, a high confinement mode plasma of >100s pulse duration with a controlled plasma density of 3.8×1019 m-3, the low H/(H+D) ratio to <10%, goal recycling coefficient <1 and core tungsten impurity concentration~10-5 was successfully achieved in EAST. These advances provide a very valuable reference for evaluating boron application in ITER and future fusion reactor devices.

[1] G.Z. Zuo. Boron coating on full metal wall in EAST for supporting ITER new baseline, 26th PSI invited talk.
[2] W. Xu, et al. Active wall conditioning through boron powder injection compatible ELM control in EAST. Nuclear Materials and Energy, 2023, 34: 101359.
[3] G. Z. Zuo, et al. Deuterium recycling and wall retention characteristics during boron powder injection in EAST. Materials Research Express. 2023;10(12):126402.

Speaker: Guizhong Zuo

14:25 Technological developments of the W7-X and JT-60SA metallic actively cooled divertor

For a successful operation of ITER and of future fusion reactors, several fusion devices are currently being exploited and upgraded. This paper focuses on the technical developments lead in JT-60SA (Japan) and Wendelstein 7-X (W7-X, Germany) fusion devices to provide support on the long pulse operation in metallic environment using operation conditions complementary to the current existing fusion devices.
The tokamak JT-60SA has been constructed in the framework of the broader approach with strong European support. After the operation of carbon divertor, a transition to a metallic device is foreseen for the “Integrated Research Phase I”. Tungsten actively cooled divertor target PFCs (W-ACD) will be installed to provide information on high-beta, inductive and non-inductive operations in metallic environment. The conceptual design of the JT-60SA W-ACD is planned to be achieved in 2026.
In parallel, Europe is also promoting studies on the stellarator concept, considered as the backup solution for the European reactor design. For this purpose, a transition of W7-X to a carbon free environment is planned. As for JT-60SA, the concept of the W7-X W-ACD with the related validated manufacturing process is planned to be qualified in 2026.
Analyzing the boundary conditions and loads related to the future W-ACDs of JT-60SA and W7-X, some similarities exist. First, these W-ACDs are planned to handle steady state thermal heat loads in the range of 10 MW/m². The current boundary conditions (cooling conditions…) and interfaces of the already existing device dictate rather strongly the design options. The armour material is planned to be pure tungsten (W) or W alloys for both fusion devices. Moreover, one of the option for these two devices is to the join the armour material to CuCrZr manufactured by additive manufacturing, ,in which the cooling circuit is inserted. To take benefit from these similarities, projects are run within the same European work package called Divertor (WPDIV). In this paper, the concepts currently investigated for the metallic targets of JT-60SA and W7-X will be detailed and the current results of the manufacturing steps will be presented.
“This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.”

Speaker: Marianne Richou

14:50 Evaluating laser ultrasound for the in-situ inspection of plasma-facing components.

In order to operate for long pulses, magnetic confinement fusion (MCF) devices are typically equipped with actively-cooled plasma-facing components (PFCs). These have been installed in several existing long-pulse MCF machines, but are expected to be more widely used in the next generation of high-powered, long-pulse devices (e.g. ITER, DEMO & commercial power plants). The circulation of fluid coolant inside PFCs introduces the risk of in-vacuum vessel loss of coolant accidents (in-VV LOCAs) should the PFCs suffer excessive in-service damage. Such accidents could have serious consequences for nuclear safety and machine availability (particularly for future MCF devices handling larger quantities of radioactive material than present or past machines), and as such every effort should be taken to prevent them. Theoretical models of PFC ageing and extrapolations from representative experiments may enable predictions of PFC service lifespan, but these will carry significant uncertainty in the case of first-of-a-kind machines operating in unexplored regimes. In the face of this uncertainty, methods to directly inspect PFCs for damage in-situ during their service lifespan could help to prevent dangerous failure accidents through the early detection and characterisation of defects. Additionally, the data gathered from regular in-situ inspections could improve scientific understanding of PFC ageing mechanisms and feed back into improved PFC designs. A suitable non-destructive inspection technology for this purpose must possess certain critical capabilities. Not only must the technique be capable of detecting and characterising the relevant types of PFC defect (e.g. cracks, thickness changes due to erosion & interfacial delaminations), but it must be remotely deployable in the extreme environment of an aged MCF vacuum vessel (i.e. compatible with ultra-high vacuum, radiation, magnetic fields etc.). Laser-based ultrasonic inspection has the potential to meet these requirements. By both generating and detecting ultrasound pulses in a target using laser beams, the advantages of ultrasound for through-thickness inspection can be leveraged without the need to make physical contact with the component or apply ultrasonic coupling fluid. In this work, the authors report on laboratory-scale experiments to assess the suitability of laser ultrasound for the in-situ inspection of PFCs. A twin-laser scanning system has been used to demonstrate non-contact ultrasonic imaging of artificial defects in a tungsten sample. Furthermore, a novel method for component thickness measurement using thermoelastic laser ultrasound excitation has been developed, and demonstrated on tungsten tiles with thicknesses of 2-10mm simulating various levels of erosion. The effect of surface finish on laser ultrasound signal quality has also been explored for tungsten tiles with treated surfaces. Finally, the authors demonstrate automated deployment of the laser ultrasound scanning system on a robotic arm, simulating practical in-situ deployment. The results of these experiments are discussed, to assess the level of capability that has been demonstrated and identify priorities for development toward the use of laser ultrasonic inspection inside future MCF devices.

Speaker: Matthew Riding (University of Strathclyde)

15:15 → 15:40 H&CD session

15:15 Experiences from long-pulse ECRH operation at W7-X

Reaching the milestone of > 1GJ of heating energy was an important demonstration of the physical and technical capabilities of the W7-X stellarator. This milestone was achieved using exclusively ECRH power, which was also a challenge for the ECRH system itself and the handling of ECRH-specific loads on the components in the vessel. This is especially true for high density operation and not fully absorbed heating with the second harmonic O-mode (O2), where the microwave stray radiation level was minimized by specially designed reflector tiles. The stray radiation level was analyzed for different operating scenarios. In addition, sensitive components had to be protected by special shielding, which could be validated in the long pulse operation scenarios.
In addition, high reliability of the multi-megawatt and multi-gyrotron system had to be achieved. In particular, improvements in power transmission and fast gyrotron control, which are reported here, led to the above mentioned success.

Speaker: Heinrich Laqua (Max-Planck-Institute for Plasma Physics, Greifswald, Germany)

15:40 → 16:10 Coffee Break

16:10 → 17:25 H&CD session

16:10 TOWARDS A LONG PULSE ECRH SYSTEM FOR WEST

The objective of WEST experiments is to master long-pulse operation (1000 s) while exposing actively cooled ITER-like tungsten divertor to power fluxes up to 10 MW/m2. To increase the margins to reach H-Mode regime and to control W-impurities in the plasma core, the WEST ECRH system is under major upgrade with the objective to reach the capabilities of injecting 3 MW during up to 1000 s at a frequency of 105 GHz.
The former ECRH system in Tore Supra was designed for shorter pulse duration and lower power per gyrotron (~210 s and 500 kW per gyrotron). Transmission lines, DC breaks, bellows, waveguides, antenna, and diagnostics were not qualified for this new level of power and duration (1 MW per line with 1000 s duration). To match these new requirements for WEST, the system has been re-designed and will be equipped with:
• A new antenna actively water-cooled (30 bar, 70°C) with a temperature monitoring of the mirrors;
• Transmission line components with dedicated cooling system with also a temperature monitoring;
• A few dedicated water cooling loops for gyrotrons (6 bar, 25°C) and the transmission lines (5 bar, 25 °C).

Diagnostics already designed for the previous system will be also used, like:
• Embedded thermocouples in the gyrotrons to protect the electron beam collector;
• RF measurement for the spectral quality at the output of the gyrotrons and a power detection of mode loss during the pulse;
• Arcs monitoring in the transmission lines.

All these diagnostics will be used in a fast safety system able to stop the RF pulse within few µs in case of issues of overheating or overpressure in the gyrotrons, transmission lines and antenna.
The paper will provide an overview of the major revisions and upgrades of the ECRH system for WEST long pulse operation in the W environment.

Speaker: Jean-Michel BERNARD (CEA)

16:25 Performance of PAM launcher in ADITYA-U tokamak

After the successful integration of passive active multijunction (PAM) launcher in ADITYA-U tokamak, the lower hybrid current drive (LHCD) experiments with PAM launcher is conducted in ADITYA-U tokamak. The PAM launcher, designed to launch up to 250kW of rf power at 3.7 GHz, for one second, was successfully installed on radial port#5 of the tokamak. It was validated for UHV compatibility (10-9 mbar) and baking tests up to ~150oC. Good coupling of PAM launcher with plasma was demonstrated both with, low (~100mW) as well as with high (~120kW) rf power. Reflection co-efficient (RC) below 5% was achieved even when PAM launcher was placed ~15mm behind limiter.

The PAM launcher is designed to launch lower hybrid waves (LHWs) having parallel refractive index (N||) centered at 2.25 when two adjacent modules are relatively phased at 180o. Experiments are carried out by varying the phase difference between the two modules from 30o to 270o. The maximum power launched in to the plasma was around ~120kW. LH power up to ~200ms could be coupled to the plasma. Initially target plasma for LH experiments was formed by using positive convertor only where loop voltage was not available beyond 65ms to sustain plasma current inductively. Toroidal magnetic field was kept at ~1.4T. In this configuration, plasma current up to ~220ms could be sustained with PAM launcher and significant current could be driven with zero loop voltage towards the end of the plasma shot. Further, to have longer Ohmic discharges, negative convertor was augmented with positive convertor, to provide more Ohmic flux, thereby providing loop voltage up to ~330ms. Plasma current up to ~430ms could be sustained with PAM launcher when target plasma was formed with negative convertor and significant plasma current could be maintained with zero loop voltage towards the end of the shot. LH experiments were carried out with both hydrogen and deuterium plasmas. Experiments suggest that low RC of PAM launcher could be maintained in the presence of ECRH pulse and it remained insensitive to plasma movements. The enhancement in the photon counts, measured by scintillation CdTe detector and the 2nd harmonic ECE signal, with the injection of lower hybrid waves, confirmed the generation of suprathermal electrons due to interaction of LHWs with plasma and eventually driving plasma current non-inductively. The spectral broadening of the pump waves represented by the pump width is also observed indicating PDI effects at the plasma edge.

In this paper, the first experimental results obtained with PAM launcher in ADITYA-U plasmas will be presented and discussed in details along with future work plans.

Speaker: Promod Kumar Sharma (Institute for Plasma Research)

16:40 Feedback control of LHW-plasma coupling for long pulse operation in EAST

Lower hybrid current drive (LHCD) has proven to be one of the most efficient methods to sustain long pulse plasma operation in tokamak. In order to sustain good LHW (lower hybrid wave) -plasma coupling required for long pulse plasma, it is the first time that the coupling feedback control is designed and realized in EAST through PID (Proportion Integration Differentiation) method by choosing RC (reflection coefficient of LHW power) as the reference for the feedback of gas puffing, including one pulse test and multi-pulse experiments.

Experiments show that the designed feedback control works correctly and keeps good LHW-plasma coupling effectively for long time, suggesting the possibility of feedback control application on LHW-plasma coupling in long pulse plasma. Furthermore, during the process of feedback control of multi-pulse SMBI (supersonic molecular beam injection), the stored energy changes from 29kJ to 58kJ, and the energy confinement factor (H89) increases from 0.98 to 1.45, implying a positive effect of coupling feedback on plasma performance. In addition, experiments between SMBI fueling and the gas-puffing fed by piezoelectric valve near the antenna are further investigated, showing that the response time on the modification of edge density with SMBI is faster than that by the piezoelectric valve and that SMBI puffing in the electric drift side of LHW antenna is a little quicker than that in the ion drift side.

Studies suggest that the feedback control through RC is effective for long pulse LHW-plasma coupling and that the gas puffing by SMBI in the electric drift side of LHW antenna offers a new and effective method to sustain good LHW coupling in steady state operation in future, encouraging the LHCD application in fusion reactor. Further optimization will be continued later.

Speaker: Bojiang Ding (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)

16:55 Lower Hybrid current drive long pulse operation state of the art on WEST

Abstract:

The objective of WEST experiments is to master long-pulse operation (1000 s) while exposing the actively cooled ITER-grade tungsten divertor to power fluxes up to 10 MW/m2. The reliability and performance of the Lower Hybrid Current Drive (LHCD) system are critical for the success of long pulse operation on WEST as it allows to drive significant non-inductive current in the plasma and reduce the magnetic flux consumption in the central solenoid [1].
The two LHCD launchers, each one powered by eight klystrons capable of delivering 600 kW during 1000 s each, can provide a total of 7 MW of power to the plasma at a frequency of 3.7 GHz. The complete system is actively cooled and each part has been qualified to support power for 1000 s. Safe plasma operation is ensured by simultaneously monitoring the infrared signals on the launcher and copper impurity density. These signals are feedback controlled in real time with the LHCD power to protect the LH launchers against excessive heat.
The LHCD system has been routinely used on WEST since 2018 with achieved performances ranging from 5.8 MW for 5 s and 3-4 MW for 364 s [2]. Continuous efforts have been made to update the tools required to maintain high level of reliability of such a complex system. In particular, a local acquisition system has been developed to operate and test all klystrons individually on water load during shutdown periods. This acquisition system is being improved further to monitor in real-time the safety-relevant parameters of the 16 klystrons.
In this contribution, we present an overview of the current status of the LHCD system on WEST, as well as the recent upgrades developed and implemented to overcome the limitations and the issues encountered on the system when both power level and pulses duration are increased.

[1] R. Nouailletas et al., this conference.
[2] T. Fonghetti et al., this conference

Speaker: Xavier Regal-Mezin (CEA, IRFM)

17:10 Experimental research of EAST start-up towards the first plasma of ITER

The experiments on plasma initiation performed in EAST with the electron cyclotron wave (ECW) pre-ionization and assisted start-up have demonstrated that ITER can produce plasma initiation in a low toroidal electric field (<0.3V/m). The parameter domain of breakdown is significantly extended towards higher prefill gas pressure. The effect of ECW injection timing, power, toroidal injection angle on breakdown were also investigated with commonly used wide null-field configuration (NFC), in which the stray poloidal magnetic field is lower than 2×10^(-3) T in most area of the vacuum vessel. The electron cyclotron heating (ECH) power threshold for breakdown in EAST is approximately 0.4 MW. In the range of ECH power tested in this work, higher ECH power is advantageous for achieving earlier and faster breakdown. During the ECW-assisted startup, the process of burn-through is prolonged by the higher pre-filled gas pressure even though it enhances the ease of breakdown. Besides the wide NFC, the newly developed mirror-like trapped particle configuration (TPC) was also tested for the ECW assisted startup. The comparison of different poloidal magnetic field configurations shows that the particle confinement and the formation of closed flux surface strongly depend on the shape, not only the strength of the magnetic field. The ECW assistance has an effect in preventing the generation of runaway electrons and improving the safety of device during start-up with the comparison of LHW assistance. The ECW assistance also exhibits a high tolerance to the impurity and thus ensures a high ramp rate of plasma current even with a high impurity level. In addition, simulations of EAST plasma initiation with the consideration of engineering design models of the PF coil, vessel and PF power systems, have produced the time-evolution of plasma current, electron density and temperature etc, suggesting a reasonable agreement with experimental measurements.

Speaker: Jinping Qian (Institute of plasma physics, Chinese academy of sciences)

Wednesday, 16 October

09:00 → 10:25 RAMI & LPO session

09:00 EU-DEMO: pulsed vs. steady-state solution

In the European roadmap to fusion energy, EU-DEMO will be the first machine with a net electricity generation and demonstrating the integration of all reactor-relevant functions, e.g. self-sustaining tritium production. Currently, the EU-DEMO design is based on the tokamak configuration, with a pulsed plasma and a discharge duration not shorter than two hours. In general, the EU approach favours lower risk solutions (although not low risk in absolute terms), in order to maximize the chances of mission success in the foreseen time. Thus, while the advantages of a steady-state tokamak operation are recognized, especially for a future commercial reactor, the pulsed solution relying on inductive plasma current drive has so far been pursued. The higher risks of fully non-inductive scenarios concern the uncertainties in the associated physics assumptions and the challenges to integrate and reliably operate the auxiliary current drive systems. The large fraction of auxiliary driven plasma current negatively impacts on the net electricity output and plant availability, or, in other words, substantial improvements w.r.t.the failure rate of H&CD systems, as well as on the wall-plug efficiency as compared to present day experiments, would be necessary to meet the EU-DEMO stakeholder requirements. Furthermore, high beta, steady-state scenarios with lower plasma current and high confinement capability often require additional plasma control strategies and actuators, e.g. for the tailoring of the safety factor profile or on the resistive wall modes (RWMs), adding further complications to the already quite challenging target of a basically disruption-free reactor. Although the central solenoid in a non-inductive machine can be significantly smaller, allowing in principle a reduction of the plasma major radius, the effective shrinking of the machine size is very much limited by other factors, like the neutron shielding and the size of the TF coils. The balance of plant solutions to cope with an intermittent thermal power generations in a pulsed device are also presented

Speaker: Mattia Siccinio (EUROfusion Consortium)

09:35 Plans to Develop Integrated Core-Edge-Wall Plasma Solutions for a Fusion Pilot Plant with DIII-D

A critical challenge for a compact fusion pilot plant is to resolve the path to a high fusion performance core with a dissipative divertor approach, finding solutions that are mutually compatible to marry them together. This must also be integrated with surrounding plasma interacting technologies and materials. An upgrade to DIII-D is proposed to close gaps on reactor physics regimes in divertor, SOL, wall, pedestal and core to test critical physics, pioneer solutions and resolve their mutual compatibility.

The key to developing an integrated solution is to raise pressure. This enables high density to be sustained at low collisionality to achieve the high dissipation divertor and at the same time a high-performance low collisionality core. On DIII-D this is achieved through a rise in shaping, current, volume and power, exploiting the natural properties of improved pedestals at high shape to close gaps and push limits. Upgraded flexible heating and current systems (electron cyclotron and neutral beam power rises) enable development of a range of pulsed and steady state core solutions at the higher densities needed for core-edge integration. Integrated modelling projects betaN’s up to 5, with unique access to low collisionality, thermalized, peeling limited reactor-like regimes, and short neutral penetration depths into the core to study relevant particle and impurity transport. The resulting increased parallel heat flux and density raise opacity and shorten mean free paths to access reactor relevant physics in the divertor.

A staged divertor program is underway to explore the development of advanced closure schemes and to isolate divertor physics mechanisms, before proceeding to a core-edge optimized configuration for integrated solutions. This will be combined with a new reactor relevant wall, where final choice of material and missions are still being discussed with the community. Also planned are tests of a further change to pumped high power closed divertor for negative triangularity operation. Passive coil runaway electron dissipation and high speed pellet disruption mitigation schemes will also be tested, while the EC power rise will be used to resolve mitigated ELM solutions in conjunction with DIII-D flexible 3D and profile control capabilities. Spin polarized fusion concept tests are also being explored.

These exciting developments will enable DIII-D to pioneer integrated core and edge solutions, their materials compatibility and required control to resolve and project the approach for the pilot. Community input is sought on how best this might complement international facilities to close gaps on fusion energy.

*Supported under DOE DE-FC02-04ER54698.

Speaker: Richard Buttery (General Atomics)

10:00 Development of Technologies for the Fusion-Fission Hybrid Systems

Environmentally acceptable nuclear energy is an important component of global energy today. However, the difficulties associated with the production of nuclear fuel, reprocessing and disposal of radioactive waste from fission reactors limit its development. The integration of technologies can ensure long-term and sustainable development of the world energy system. One of the most promising ways to solve these problems are fusion neutron sources (FNS) using the fusion reaction of heavy hydrogen isotopes: D and T. The relevance of the development of a hybrid fusion-fission reactor system lies in the possibility of using thermonuclear neutron fluxes to produce fuel nuclides, for example U-233.
The project of creating a hybrid installation is aimed at achieving long-pulse operating modes with a load on the first wall of up to ~ 0.2 MW/m2 of neutrons with an energy of 14 MeV and neutron fluence during the life cycle of the installation of ~ 2 (MW x year)/m2 (10 years of continuous operation).
Within the framework of hybrid technologies being developed by the Kurchatov Institute Research Center, the current tasks are: development of a project for an experimental compact neutron source (CNS), including the definition of technological schemes of the main installation systems, as well as the development of prototypes of various variants of the nuclear blanket, demonstration of the possibility of obtaining the isotope U-233 from the natural isotope Th-232. In connection with the expected results of experiments on the tokamak T-15MD, it seems advisable to consider the possibility of creating the CNS with the scale of this installation.
The design goals for the period 2021-2024 are focused on the development of new plasma modeling tools and operational scenarios, improving the performance of auxiliary systems and integrating devices that implement upgraded and new technical solutions. These include the first wall, divertor, NBI, fuel filling and pumping, ECR heating and current drive, heat transfer, fusion and nuclear fuel cycles, lithium technologies.
The conducted economic assessments have shown that the cost of fissile nuclides obtained because of processing solid-state spent nuclear fuel and fuel from the molten salt layer is comparable. However, it should be borne in mind that in a hybrid system with suppressed fission, fewer minor actinides and radioactive waste are formed than in nuclear reactors. Therefore, it seems expectable that the cost of fissionable nuclides produced using a hybrid blanket can potentially increase economic efficiency.
A comparative analysis of the radiation damage to the first wall of the hybrid installation was carried out when replacing the beryllium coating with a tungsten one. The results showed that tungsten is the best material for the first wall. However, it is necessary to take into account its increased fragility with a significant time exposure to high-energy neutrons.

Speaker: Yury Shpanskiy (NRC "Kurchatov Institute")

10:25 → 10:55 Coffee Break

10:55 → 12:40 RAMI session

10:55 Nuclear analysis towards a spherical tokamak for energy production: analysis approaches and measurement experience derived from JET DT operations

Predictions of fusion’s radiation fields and associated nuclear quantities are fundamentally needed to serve a range of performance, safety and regulatory criteria - at all stages of the plant lifecycle, ranging from conceptual design to decommissioning. Radiation fields comprise the spatial, energy and temporal distribution of neutron and photon fluxes throughout the fusion device in various phases of operation. These also extend to several associated derived quantities, for example, nuclear heating, damage, dose and radionuclide production rates in those materials comprising the technology. Nuclear analysis, encompassing radiation transport methods, inventory codes, nuclear data, geometric-material models, and computational analyses, constitutes the framework for predicting these responses.

Confidence in neutronics predictions for a fusion power plant depends primarily on: i) the specific knowledge of design and operations to be analysed; ii) the quality of the nuclear data and validation base, rooted in current and historical nuclear experiments; and iii) the validity of computational methods. This work examines UKAEA’s neutronics methodologies tailored to a spherical tokamak based on a design from the STEP programme. A parametric approach is employed to assess on-load neutron and photon fields as well as residual photon radiation fields and waste arisings through pulsed and steady-state operating scenarios.

This work discusses underpinning experiments through post-irradiation nuclear measurements derived from unique deuterium-tritium operations at JET, supported by international verification and validation efforts, including those within the SINBAD and JADE frameworks, and the IAEA FENDL nuclear data library.

Speaker: Lee Packer (UKAEA)

11:30 WEST - LONG PULSE OPERATION IN A TOKAMAK

Since 2016, the WEST tokamak has brilliantly demonstrated its capability to perform long plasma discharges in a fully metallic environment.
WEST, qualified as a long pulse machine, presents the specific features of a permanent magnetic field produced by 18 Superconducting Toroidal Field coils cooled at 1.8K thanks to its liquid helium cryogenic system, the entire plasma facing components including the tungsten ITER grade divertor actively cooled by a pressurized water loop. WEST offers a representative and complete environment for next step devices such as ITER, JT-60SA and DEMO.
The last two experimental campaigns in 2023 and 2024 are crowned by experimental results and performance continuously improved. The machine availability reached 80% for more than 8 months. The integrated plasma operation exceeded 5 hours for each campaign and energy handling capabilities was demonstrated with 80 GJ of cumulated energy injected with Lower Hybrid Current drive (LHCD) and Ion Cyclotron Resonance Heating (ICRH) systems. In addition, two records of plasma duration of 101s in 2023 and 364s in 2024 were obtained.
Operation of long pulse in a tokamak requires on one hand, machine availability achieved by maintenance and evolutions on sub-systems, technical management of unplanned interventions and operation coordination. On the other hand, the capability to ensure protection of Plasma Facing Components (PFC) during long steady state discharges is essential. The paper presents selected highlights about the strategy put in place on the WEST Tokamak.
A key point in improving the responsiveness of technical interventions has been the automatization of water leak detection through the evolution of the complex water network configuration. As this point is crucial for the future machines, advanced techniques are developed, such as infrared water leak detection and remote sniffing inside vacuum vessel.
Based on systematic electronic data logging and manual record of events, a complete technical performance study has been performed covering the two last campaigns. The analysis focus on tokamak reliability, operation starting time, pulse rate, pulse rating and actions to increase available time for experiments.
For PFC protection, a remarkable set of infrared diagnostics, covering 52% of the first wall of the vacuum vessel, provides a large thermal mapping, identifies thresholds and deliver an alarm to the plasma control system to tune the heating power in order to prevent over-heating and continue plasma discharge.
The upcoming campaign is now preparing with the implementation of the new ECRH heating system and the goal of a plasma duration of 1000s.

Speaker: Mrs Valerie LAMAISON (CEA Cadarache)

11:55 WEST: the challenges of reaching 100% of actively cooled components for long pulse operation

The WEST (Tungsten Environment in Steady-state Tokamak) project in France is crucial for reaching sustainable fusion energy by simulating reactor-like conditions. WEST provides a unique facility to integrated and test technologies for Long Pulse Operations.

Previous operation on Tore Supra, and numerous studies showed that all components receiving convective heat flux from particles and thermal radiations from the plasma must be water-cooled. Key components such as the divertors and limiters were identified for their high thermal load. Their design has been studied for many projects and is well documented. The integration process involved rigorous design and manufacturing standards, particularly focusing on over 2000 Copper/Stainless Steel junctions critical for preventing water leaks during operation.

Additionally dedicated models were developed to compute thermal loads on all components within WEST Vacuum Vessel (VV), including the VV inner shell, ports walls, diagnostics… It was shown that depending on the plasma scenario and the reflectivity hypothesis of the surfaces, temperatures can significantly vary, forcing the design teams to take into account conservative envelop cases. Dedicated water cooled protection panels were integrated to shield the larger surfaces of the VV. Actively cooled protections on some of the most loaded diagnostics inside the ports were developed. These upgrades were installed during several annual shutdowns. The current WEST configuration includes more than 98% of the VV surface with actively cooled components. Innovative technologies are not indispensable, however all components must be manufactured and installed under high quality requirements, and stringent tightness control methods to prevent any risk of leak over the years from the numerous welds and hundreds of meters of cooling channels.

However, some areas were too complex to cool down, for two main reasons: either it was too difficult to integrate water cooled components, or it was too complex to route water pipes to feed these components and to perform all welds with the required quality. Those aspects cannot be overlooked since they can severely affect the operational domain for long pulse operations. The remaining inertial structures are monitored with thermocouples and IR diagnostics.

Flowmeter and thermocouples are installed in all water loops and a dedicated calorimetry diagnostic was developed. The energy balance is closed with an imbalance of about 10% of the total injected energy for most of the campaigns. Those diagnostics are mandatory to monitor the behaviour of poorly cooled components, to check that the cooling remains efficient during long pulse operation (no flow perturbation or critical flux…) and allow assessing the lifetime of inertial components.

Yet, degassing occurred during long pulse operation, leading to an increase of plasma density and eventually disruptions. This is likely caused by overheating of surfaces. Indeed, thermocouples show that the temperature is steeply rising on some areas around the WEST upper divertor, reaching values above 300°C after 300s pulses and 1GJ injected. Some inertial areas not controlled by thermocouple temperature measurement may also have caused this outgassing.

The path to high power continuous operation in WEST still requires improvements of protection of its internal components.

Speaker: Lionel MEUNIER (CEA)

12:10 ENEA Fusion Components Failure Rate Database, status and evolution

Nuclear fusion is expected to be available as energy technology source of energy from mid-50s of present century. One key aspect for the success of such technology will be the reliability growth implemented in the design since the early phases. In fact, nuclear fusion plants are expected to exploit many fusion specific components with relatively low technology readiness level, or in some cases exploit components with mature technology from fission plants but within different and poorly explored operating domain window (e.g., in terms of loads, operation regimes, etc.). An important source of uncertainty in reliability assessment then resides in the failure and repair model definition for fusion plant components.
In order to limit such uncertainty, ENEA Fusion Components Failure Rate Database was developed to collect failure and repair screening data suitable for fusion systems reliability analyses. The records of the database were validated by field expert collaborative effort in the context of International Energy Agency program and mainly consist of two categories: i) nuclear fission data recommended values to use for fusion applications ii) data estimated from nuclear fusion research facility operating experience.
First a review of available data, database organization status is presented. Then database evolution is described in terms of recently added failure and repair data records derived from engineering selection, specific component selection datasheet or failure modeling.

Speaker: Dr Danilo Nicola Dongiovanni (ENEA, NUClear Department, Frascati Research Center)

12:25 Enhancing Efficiency and Reliability of Long-Pulse Tokamak DC Transmission Systems through Innovative Electromagnetic Topology Optimization

Abstract—The pursuit of fusion energy as a sustainable and clean power source has spurred the development of advanced fusion devices, exemplified by the China Fusion Engineering Experimental Reactor (CFETR) and the Burning plasma Experimental Superconducting Tokamak (BEST). These devices aim for long-pulse steady-state operation, presenting significant challenges for their DC transmission systems. This study focuses on optimizing the electromagnetic topology and parasitic parameters of the Tokamak DC transmission system for long-pulse steady-state operation. The primary challenge is to enhance system efficiency and reliability. Through thorough exploration of electromagnetic challenges, innovative strategies are proposed. Utilizing numerical simulations and experimental validation, this study reveals critical electromagnetic characteristics and quantifies the impact of parasitic parameters. The core innovation lies in developing tailored optimization schemes for long-pulse operation conditions, addressing unique challenges such as high-parameter voltage withstand levels (up to 36 kV) and rated operating currents (up to 55 kA). By integrating theoretical analysis, finite element methods (FEM), and experimental validation, this study explores the electromagnetic characteristics of the DC transmission system, including magnetic field distribution and current density. Additionally, parasitic parameters such as stray inductance (≤0.2μH/m) and resistance (≤2μΩ/m, including two-pole conductors) are analyzed and quantified for their impact on system performance. Optimizing system performance is crucial to ensure reliability under long-pulse operation conditions. The research findings demonstrate that optimizing electromagnetic topology, improving electric field distribution, and optimizing parasitic parameters significantly enhance the performance of long-pulse steady-state operation Tokamak DC transmission systems, providing crucial support for the realization of sustainable fusion energy.
Keywords- Fusion energy, Tokamak, DC transmission system, Electromagnetic topology, Parasitic parameters, Long-pulse, Optimization, Innovation, Numerical simulation

Speaker: Zhengyi Huang (University of Science and Technology of China, Hefei 230026, China. Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)

12:40 → 14:10 Lunch Break

14:10 → 15:30 H&CD session

14:10 Towards long pulse operation of N-NBI ion sources

Neutral Beam Injection (NBI) requires high particle energies if one of its aims is to contribute to current drive in large fusion tokamaks. For example, 1 MeV D is foreseen for the ITER NBI. At such energies, the NBI must be based on a source of negative hydrogen ions (N-NBI) due to their higher neutralization efficiency of up to 60% in a gas neutralizer. Negative hydrogen ions are produced on low-work function surfaces, for which caesium is evaporated continuously into the ion source. The strong technological development resulted in an ion source that operates technically reliably and is in principle capable of running continuously (RF plasma generation, RF coupling, high voltage for extraction and acceleration, cooling etc.), where the only technical limit for the operating time is the vacuum pumping capacity. However, the strong dynamics of the Cs layers caused by the plasma-surface interaction creates a steadily increasing amount of inevitably co-extracted electrons, limiting the pulse duration at present. The vacuum pumping and further aspects regarding the neutralizer, beam duct components, etc. are discussed in [1].

The N-NBI test facilities BATMAN Upgrade and ELISE (1/8 and 1/2 size of the ITER NBI source, respectively) contribute to the development program of the ITER NBI; while full size prototype sources are hosted by Consorzio RFX at the Neutral Beam Test Facility (Padova, Italy). Conditioning recipes and further measures (e.g. biased surfaces close to the extraction system) have been developed to stabilize and/or reduce the current of co-extracted electrons. These optimizations resulted for the first time in almost 90% of the targeted extracted negative ion current (30 A) reproducibly achievable in 600 s hydrogen pulses at ELISE, demonstrating that the requirements for the first operational phase of the ITER NBI are in reach for large scale N-NBI sources. Deuterium operation remains more challenging, since the co-extracted electrons increase more strongly during a pulse. A newly developed measurement of the work function proved that the work function degrades after long pulses. Simulations give hints that the limited flux of Cs onto the surface during the plasma discharge causes the degradation. In order to increase the flux of neutral Cs, an alternative Cs evaporation concept (“Cs shower”) is tested at BATMAN Upgrade. With the Cs shower, a steady state with extremely stable performance has been reached in long deuterium pulses for the first time in a caesiated negative ion source. Further improvements of the concept are required in order to make it applicable to large sources.

This contribution reports on the significant progress of N-NBI ion sources towards long pulse operation. Results from plasma and beam diagnostics are presented and possible solutions to bring large ion sources into a steady state are discussed.

[1] C. Hopf et al., “Towards long-pulse and continuous positive-ion-based neutral beam injection”, this meeting.

Speaker: Christian Wimmer (Max-Planck-Institut f. Plasmaphysik)

14:35 Investigations of the RF coupling efficiency in the MINION device

In order to secure the Long-Pulse operation of ITER reactor it is foreseen to use two Heating Neutral Beam Injectors (HNB), each one expected to inject into the plasma a beam composed of deuterium atoms accelerated up to 1 MeV energy, delivering a power of up to 16.5MW for a beam pulse length up to 3600 s. Since these operating conditions have never been reached jointly before a dedicated Neutral Beam Test Facility (NBTF) is hosted at the Consorzio RFX (Padua, Italy). The goal of this experiment is to generate fast neutral beams by accelerating and neutralizing negative ions, produced in a RF inductively-coupled plasma (ICP).
The prototype negative ion source used for this purpose (SPIDER), consists of 8 driver volumes and an expansion chamber containing static magnetic filter. Recent results from SPIDER indicate plasma non-uniformities, which affect the properties of the beam leading to unexpected large beam divergence (> 7 mrad). In order to understand the reasons for this beam behaviour and optimize the long pulse performance of the source, one separate RF driver has been installed in a more accessible and flexible experimental setup (MINION).
This contribution presents the results of numerical studies of plasma parameters in the new experiment MINION carried out in order to optimize the inductive coupling between the radio-frequency (RF) active currents and the source plasma.
Analyses are done by means of the numerical code FSFS2D which gives self-consistent two-dimensional description of the source, including neutral gas flow, plasma chemistry, RF coupling in the source driver and plasma transport through the magnetic filter.
The paper outlines the main assumptions of the RF model and necessary code developments to simulate the currents in the source induced by the RF coils. The RF electrical field is split in a plasma and a vacuum part, E = E^p + E^V. This simplifies the boundary conditions for E^p; whereas E^V is obtained from a theoretical formulation for the field generated by a current loop; such fast flexible algorithm provides converging RF field solution after a few iterations, even with high plasma conductivity.
RF-plasma coupling determines the characteristics of the plasma in the source and therefore studies are carried out to optimize the power transfer efficiency in terms of the main external controllers of the driver like the power output from the generator (which determines the value of the RF-coil current) and the gas pressure in the chamber of the driver. Special attention is put on the effect of the different magnetic fields (cusp magnets, magnetic filter, RF magnetic field) on the coupling efficiency. It is found that a reasonable range of RF currents in the coils is obtained only under assumption of strong reduction of plasma conductivity by the magnetic fields.

Speaker: Prof. Roman Zagorski (National Centre for Nuclear Research (NCBJ), 05-400 Otwock, Poland)

14:50 Towards long-pulse and continuous positive-ion-based neutral beam injection

While Neutral Beam Injection (NBI) with beam energies in excess of about 100 keV/amu (e.g. on ITER, JT-60SA, DTT) requires sources for negative ions (“N-NBI”), positive-ion-based (“P-NBI”) systems have attracted renewed interest for smaller long-pulse fusion devices such as volumetric neutron sources (VNS). In these devices, like for instance the tokamak-based VNS currently studied by EUROfusion [1] or the high-field mirror BEAM proposed by Realta Fusion Inc. [2], D—T fusion is chiefly driven by fast ions from NBI. The several hours long pulses with high duty cycle or even continuous operation that these devices aim at sets unprecedented requirements for NBI.

The experience with inductively coupled radio frequency (RF) ion sources [3, 4] shows that they are already very reliable and, particularly for P-NBI, basically maintenance-free. Technically demanding solutions required for continuous operation, such as active water cooling of the sources’ Faraday screens, have also been successfully implemented in the N-NBI derivatives of the RF ion sources. However, the gap between the maximum accumulated beam-on times per source on today’s facilities and those foreseen at VNS devices still remains huge. This also means that so far irrelevant phenomena, such as erosion of the high heat flux components by physical sputtering, may very well determine the maintenance schedule of the beamline.

Another key aspect is pumping. NBI requires high feed gas flows to the ion source and neutraliser. In order to keep losses of the neutralised beam by re-ionization low in the beamline, this gas must be efficiently pumped by large area getter pumps such as cryo pumps. These pumps require recurring regeneration when the limit of the adsorbed hydrogen inventory is reached. The limit is usually determined by safety considerations. Continuous NBI therefore requires concepts of accommodating regeneration without interrupting injection, e.g. by cyclically operating N – 1 out of N beamlines while one beamline regenerates. This in turn requires improvements in gas consumption and pump regeneration time.

Continuous operation also increases the relevance of NBI’s overall energy efficiency, as NBI will be a dominant power consumer of a VNS. Developing concepts to improve the beamline’s wall-plug efficiency, such as residual ion beam energy recovery, to maturity deserves increased attention.

The paper will give an overview of the challenges, strategies and development needs to make P-NBI meet the demands of continuously operated, NBI-reliant fusion devices. Many of the aspects are relevant to N-NBI as well.

[1] G. Federici, Nucl. Fusion 63 (2023) 125002
[2] C.B. Forest, IAEA FEC (2023)
[3] E. Speth et al., Fus. Eng. Des. 46 (1999) 383
[4] B. Heinemann et al., New J. Phys. 19 (2017) 015001

Speaker: Christian Hopf (Max-Planck-Institut für Plasmaphysik)

15:15 NUMERICAL STUDY AND OPTIMIZATION OF NON-INDUCTIVE CURRENT DRIVE EFFICIENCY IN FNS TOKAMAK PLASMAS

Tokamak-based fusion neutron sources (FNS) can effectively address the fuel problems of nuclear energy and future fusion power plants operation [1]. FNS primary mission is testing technological systems being exposed to intense neutron flux and evaluation of relevant effects in materials and structural components designed for future nuclear reactors, fusion-fission and pure thermonuclear devices. Hybrid facilities based on FNS can provide invaluable experimental data on nuclear waste management, nuclear fuel production and other technologies for future hybrid reactors. Ultimately, a steady state tokamak-based neutron source with intensity level $10^{18} - 10^{19}$ neutrons per second can become a cost effective solution when compared to other neutron generators.

At present, two major design options of FNS are considered and still a subject for comparison: a conventionally shaped tokamak DEMO-FNS with a moderate aspect ratio ($R_0 = 3.2 m, a = 1 m, k = 2.1, B_0 = 5 T, P_f = 400 MW, Q = 1$), and a strongly shaped spherical device FNS-ST ($R_0 = 0.5 m, a = 0.3 m, k = 2.75, B_0 = 1.5 T, P_f = 2 MW, Q = 0.2$). Both tokamak designs assume steady state operation sustained by bootstrap and auxiliary current drive (CD). Neutral beam injection (NBI) is chosen as a main source of energetic particles for plasma fuelling, heating, torque, and as the most efficient current driver (NBCD), with high efficiency confirmed theoretically [2] and experimentally [3].
In theory, NBCD efficiency ($η = I_{CD} \lt n_e\gt R / P_{NB} / 10^{20}$) is much higher when compared to other auxiliary CD sources [4, 5]. Our numerical analysis predicts η = 0.2-0.5 can be achieved in optimized scenarios for both FNS devices. In reality, NBCD values strongly depend on the magnetic field shape, NB size and injection geometry, on plasma target density and temperature. The target values of non-inductive current drive are: ~5 MA (DEMO-FNS, $P_{NB}$ = 30 MW), and ~1.5-2 MA (FNS-ST, $P_{NB}$ = 6-10 MW). NBI parameters are optimized either to maximize NBCD efficiency in a given plasma or, alternatively, to obtain a hollow current profile favourable for steady-state operation. NBCD and bootstrap profiles for expected range of DEMO-FNS and FNS-ST scenarios are presented in this contribution.

A simple numerical model [6] is used for beam-driven effects analysis in tokamak plasmas. It incorporates the beam detailed structure with account of NB producing technology and the ions tracking in the equilibrium magnetic field. MHD equilibria are reconstructed by the analytic solution of Grad-Shafranov equation [7]; the approach allows the study of plasma shaping effects through configurations of interest - from conventional to spherical tokamaks. NB fast ion distributions and resulting NBCD profiles are obtained by fast ion tracking in magnetic field. For both FNS devices, non-inductive CD is expected to reach the target values. The results prove NBI is a plausible solution for accessing the long pulse plasma operation and control in FNS, if NBI parameters and injection geometry are chosen appropriately.

[1] Shpansky Y.S. and DEMO-FNS Project Team, 2019 Nucl. Fusion 59 076014

Speaker: Mrs Eugenia Dlougach (NRC KI)

15:30 → 16:00 Coffee Break

16:00 → 16:50 H&CD session

16:00 NBI performance for long pulse operation in KSTAR

Neutral beam injector is one of the essential devices for KSTAR long pulse operation. It contributes to sustain the non-inductive plasma current in long pulse tokamak operation as well as to obtain the high ion temperature operation mode in KSTAR. In a viewpoint of long pulse superconducting tokamak, it has been playing a very challenging pioneering work for future tokamak reactor since its first operation in 2008. In this talk, KSTAR NBI performances are discussed in terms of beam power stability in long operation, beam species and divergence on the plasma performances, and dependence on low and high Z PFC wall.

Speaker: Jonggu Kwak (KFE)

16:25 Optimizing fast-ion confinement in NBI plasma for long-pulse operation of EAST

It has been observed that lost fast ions in NBI plasma will hit the first wall of the device to affect the long pulse steady-state operation. The fast-ion loss mechanisms include prompt loss, ripple loss and resonant loss due to MHD instabilities[1]. Also, MHD instabilities are closely related to fast-ion beta βf [2-3]. In our research, we focused on the simulation and experimental investigation about fast-ion confinement by optimizing plasma shape to reduce prompt loss and ripple loss and adjusting plasma parameters to avoid MHD instabilities. Firstly, the influence of gapout in plasma shape on fast-ion confinement is analyzed. The simulation and experimental results show that larger gapout helps to improve fast-ion confinement due to the beam deposition moving inward, which leads to less trapped particles. Also,the prompt loss and ripple loss of fast ions are reduced. Meanwhile, the larger gapout leads to bigger shafranov shift and better confinement (higher βp and H98). Secondly, the fast-ion beta and gradient variation with beam energy and electron density have been achieved when Ip~500kA, Bt~-2.5T, q95~5.3. Simulation results show that average fast-ion beta increases with beam energy ENBI and decreases with electron density due to longer slow time. When Bt~1.6T and q95~4.4, the average fast-ion beta is about doubled. The gradient of fast ion profile is larger at lower density, which leads to the appearance of TAE and EP in EAST NBI experiments [3]. Therefore, in order to avoid the abnormal transport of fast ions caused by MHD instability, the background plasma with high Bt and high density should be selected in the long pulse experiment with NBI. In the end, the longest about 60s MHD quiescent NBI plasma has been achieved based on above optimized configuration and plasma parameters with Bt~-2.4T, fGW~0.65.

[1] N. N. Gorelenkov, et al., Nuclear Fusion,54,125001(2014).
[2] C. T. Holcomb, et al., Phys Plasmas,22(2015).
[3] L. Xu, et al., Nuclear Fusion,61, 076005(2021).

Speaker: Jinfang Wang (Institute of Plasma Physics Chinese Academy of Sciences)

16:40 High-power and long-pulse operation of ICRH system in EAST tokamak

Ion cyclotron resonance heating (ICRH) has been a dependable tool for sturdy and long pulse plasma heating with high RF power of several megawatts. However, low ICRH antenna coupling efficiency, high temperature of antenna limiter and Faraday Screen (FS) and MHD instabilities have limited high-power and long-pulse operation of the system. To increase ICRH antenna coupling efficiency and decrease the voltage in transmission line, a new kind of ICRH antenna with smaller k_(∕∕) had been designed and operated[1]. The coupling loading of the new ICRH antenna is ∼2–3 times greater than old ICRH antenna and the efficient ion and electron heating had been achieved. To operate the ICRH antenna in a better heating state and produce relatively low impurities, the antenna strap probe based diagnostic system had been designed to monitor antenna phasing, and the power and phase feedback control system had been used to control antenna phasing[2, 3]. To decrease the heat load in the front face of the ICRH antenna, new kind of antenna limiter had been designed with CFC material, which has better water cooling as the thickness of the antenna limiter with CFC material is thinner than before. And this limiter design allows the antenna straps closer to the plasma, which also could improve the antenna coupling efficiency. New kind of FS with optimized cooling channels had been designed and used, which could make the temperature of ICRH antenna FS lower than 400° in the pulse of 200 s and the power of 1 MW operation [4]. In order to make the ICRH system working sturdy during the confined mode transition and edge localized modes activity, the load tolerant matching network have been designed and operated. By keeping a low reflection ratio in the network for a wide range of resistance, this matching network could allow sturdy high-power and long-pulse operations without fast impedance matching [5, 6]. For long-pulse and high-power operation, we had achieved 1.5 MW/41 s and 1.1 MW/61 s with one ICRH antenna in the high poloidal beta plasma discharge. ICRH power up to 1.8 MW is routinely coupled to the plasma for pulse lengths up to 21 s and 0.6 MW for 300 seconds.
References
1. Zhang, X.J., et al., First experimental results with new ICRF antenna in EAST. Nuclear Fusion, 2022. 62(8).
2. Liu, L.N., et al., Measurement of the ICRH antenna phasing using antenna strap probe based diagnostic system in EAST tokamak. Nuclear Engineering and Technology, 2022. 54(10): p. 3614-3619.
3. Liu, L.N., et al., Design of power and phase feedback control system for ion cyclotron resonance heating in the Experimental Advanced Superconducting Tokamak. Nuclear Engineering and Technology, 2024. 56(1): p. 216-221.
4. Liang, Q.C., Liu, L.N., et al., Nuclear Engineering and Technology, 2023. 55(7): p. 2621-2627.
5. Liu, L.N., et al., Impedance matching system using triple liquid stub tuners for high-power ion cyclotron resonance heating in EAST tokamak. Review of Scientific Instruments, 2022. 93(4): p. 043506.
6. Liu et al 2024 Nucl. Fusion https://doi.org/10.1088/1741-4326/ad4048

Speaker: Lunan Liu (ASIPP)

Thursday, 17 October

09:00 → 10:05 LPO & Control session

09:00 Divertor-safe Nonlinear Burn Control in Long-Pulse Reactor Operation

The control of the core-plasma kinetic state determining the overall fusion power, usually referred to as burn control, arises as one of the most fundamental problems in nuclear fusion. Feedback control of the burn condition will be necessary to avoid undesirable transient performance, to respond to changes in plasma confinement, impurity content, or operation conditions, which could significantly alter the plasma burn, and to avoid potentially disruptive plasma conditions due to thermal instabilities. In order to keep the operating point far from stability, controllability and safety boundaries, and therefore avoid disruptions and protect the device while regulating the fusion power, a well-design burn controller should be able not only to achieve tight regulation around a desired operating point by rejecting perturbations in temperature and density but also to drive the plasma from one operating point to another during the burning-plasma phase, to access to and exit from the burning-plasma phase, to incorporate operational constraints arising from temperature and flux limits imposed by the plasma facing components, and to handle the nonlinear coupling with other competing controllers using shared actuators.

These control objectives demand not to neglect the nonlinear dynamics of the plasma during the control synthesis process. In order to overcome the operability limits imposed by the linearization of the burn dynamics, nonlinear techniques for burn control have been proposed to account for the non-local character of the dynamics. Control designs incorporating the nonlinear dynamics of the plasma have shown higher levels of performance, stability, and robustness against model uncertainties. These controllers utilize several actuators simultaneously, using auxiliary power modulation to prevent quenching, impurity injection (increase of radiation losses), confinement degradation (reduction of plasma energy) and isotopic fuel tailoring (decrease of fusion power density) to cool down the plasma and stop thermal excursions, and fueling modulation to regulate the density. Since the primary goal of a burn controller is usually to regulate volume-averaged properties of the plasma (e.g., fusion power or Q), zero-dimensional models are appropriate for control synthesis. However, if the goals were different for different regions of the plasma (e.g., core vs. edge), dynamic models capturing the spatial dependence would be needed for control synthesis. In any case, control testing in one-dimensional simulations would be critical before implementation.

A burn-control solution with the capabilities described above will demand not only the design of nonlinear robust/adaptive control algorithms but other key components such as: i- state observers to estimate in real time from limited-in-number and noisy diagnostics the plasma properties needed for control; ii- online optimizers to determine in real time the references needed by the controller given the desired operating point and the constraints imposed by the core-edge dynamics; iii actuator allocation algorithms to map in real time the heating and fueling requests by the controller to the available actuators while taking into account actuator dynamics and time-varying model uncertainties in the mapping. An overview of the state of the art and open challenges in burn control for long-pulse reactor operation will be provided.

Speaker: Eugenio Schuster (Lehigh University)

09:35 Control Advancements Supporting Long-Pulse Operation on KSTAR

Control developments, including porting a Proximity Controller [1] to aid in continuous disruption prevention and upgrading the existing Real-Time Feed-Forward algorithm [2] for robust and reliable shape control during long-pulse sustainment and ramp-down, have been implemented to support long-pulse scenarios on KSTAR with its new tungsten divertor. The proximity controller calculates the proximity to known stability limits and guides the plasma state away from those limits. First application of the Proximity Control is for robust vertical displacement event avoidance, with additional use cases to follow. The Real-Time Feed-Forward algorithm updates the feed-forward coil current targets in real-time based on changing plasma conditions, reducing the burden on the shape feedback systems by accounting for accumulated inaccuracies in projected plasma performance. Upgrades to the real-time feed-forward will expand on this work to handle evolving resistivity and target shapes, as well as to aid both scheduled and un-scheduled ramp-downs.
This material is based upon work supported by the US Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the KSTAR facility in Daejeon, Korea, under Award DE-SC0023399. This work was supported by Korean Ministry of Science and ICT under KFE R&D Programs of “KSTAR Experimental Collaboration and Fusion Plasma Research (Grant no. KFE-EN2301-14)”.
[1] J.L. Barr et al, Nuclear Fusion, 61, 126019 (2021)
[2] M. Walker et al, IEEE Conference on Control Applications, 605 (2016)

Speaker: Nicholas Eidietis (General Atomics)

09:50 Adaptive energy-sensitive x-ray cameras for the record long-pulse at WEST

The WEST superconducting tokamak features a full tungsten environment and is equipped with an actively cooled ITER-grade divertor providing valuable input for future operation at ITER. Versatile multi-energy soft and hard X-ray pinhole cameras have been developed, calibrated, deployed and operated for long-pulse plasmas at WEST. These cutting-edge instruments, serving as novel enabling technologies, have facilitated investigations into a wide array of phenomena including particle, impurity, and thermal transport, heating and RF current-drive mechanisms, equilibria, MHD physics, and the diagnosis of non-Maxwellian effects such as runaway electrons (RE). This innovative imaging diagnostic leverages a pixelated x-ray detector capable of independently adjusting the lower energy threshold for photon detection on each pixel. The primary detector employed is a PILATUS3 100K, equipped with 0.45 mm Si and 1.0 mm CdTe sensors, enabling sensitivity to photon energies ranging from 1.6 to 30 keV and 20 to 200 keV, respectively. Through meticulous trimming and calibration of the lower energy thresholds, our team has successfully mitigated contributions from radiative recombination and line emissions from medium to high-Z impurities like tungsten. Central electron temperature values are thus derived by modeling the slope of continuum radiation, extracted from ratios of inverted radial emissivity profiles across multiple energy ranges, without relying on a-priori assumptions of plasma profiles, magnetic field reconstructions, high-density limitations, or shot-to-shot reproducibility. Recent breakthroughs include the temporal evolution measurement of central temperature in quasi non-inductive scenarios during the C9 campaign at WEST, encompassing long-pulse L-modes (H98y,2 ~ 1.0) with a stationary central electron temperature of ~4 keV for up to 6 minutes with a total injected energy of up to 1.14 GJ. Time-intensive computer calculations also confirmed good agreement between relevant High Fidelity Plasma Simulator (HFPS) and the measurements indicating a flat central electron temperature of about 4 keV. Novel calculations of the local radiated power density profiles from photon-counting measurements in several energy bands are also now feasible. The possibility of measuring steady state tungsten transport in long-pulse scenarios – without the need of perturbative experiments - will also be discussed for the first time.

Speaker: Luis F. Delgado-Aparicio (Princeton Plasma Physics Laboratory)

10:05 → 10:35 Coffee Break

10:35 → 12:15 LPO & Control session

10:35 Long pulse operation in fully metallic tokamak WEST: control and scenario development

WEST is a tungsten tokamak designed for long pulse operation with actively cooled components. The main missions of WEST include high fluence plasma divertor exposure and the demonstration of long pulse H-mode in a full tungsten environment. Achieving long duration and high performance plasma discharges while ensuring the protection of the machine requires features such as specific controllers to keep the plasma in steady-state for several minutes and real-time machine protection algorithms to keep the plasma within operating limits.
One of the key parameters for long duration plasma is the flux consumption, which directly determines the pulse duration. To achieve this, simultaneous control of the loop voltage and the plasma current is required. The control scheme uses the voltage applied to the central solenoid and the power of the lower hybrid current drive system as actuators. These controllers have been developed and tested on the WEST flight simulator before being implemented in the Plasma Control System (PCS). A final tuning of the control parameters has been performed on only a few pulses.
Discharges of several minutes lead to triggering low dynamics phenomena such as Plasma Facing Component (PFC) temperature rise, which need to be monitored to avoid damage to the machine. Based on the infrared system measurements, the PCS decreases the heating power when surface temperatures of critical in-vessel components exceed pre-defined thresholds. If a second threshold is reached, a hard stop of the plasma is triggered.
In addition to these efficient controllers, a robust scenario has been developed. The LH power and plasma density ramp-up are set by using a scaling law to avoid ripple losses and tungsten radiation (leading to a plasma collapse). Then a slow ramp-down of the loop voltage on several tens of seconds is performed to avoid MHD and to reach a final value of 3-4 mV. Despite these strategies, unexpected events like Unidentified Flying Objects (UFOs) or long time-scale outgassing (>300 s) of in-vessel component located further away from the plasma heat load have required several repetitions to reach the current WEST plasma duration record of 6 minutes and 4 seconds with 1.15 GJ of injected/extracted energy.
The generation of UFOs during operation of a full metallic tokamak and the outgassing of remote in-vessel components are some of the main lessons learned during the recent WEST campaigns. To solve the UFO issue, the use of nitrogen injection to trigger X-point radiation seems promising. In this case, feedback control is required to maintain partial detachment without leading to a collapse of the plasma.

Speaker: Rémy Nouailletas (CEA)

11:00 Long-pulse no-ELM H-mode operation with feedback-controlled detachment and H98,y2~1.1 under boronized metal wall in EAST

Metallic impurities induced by divertor heat load and ELM filaments are the main challenges in achieving long-pulse high-performance H-mode operations. Long-pulse (>50s) no-ELM H-mode discharges with feedback-controlled divertor detachment by nitrogen seeding from outer horizontal target plate of the lower divertor have been achieved under boronized metal wall in the EAST tokamak. Figure 1 shows such a discharge with $n_{el}$~4$\times$$10^{19}$$m^{-3}$, $I_p$=400kA, $B_t$=2.45T, $q_{95}$~6.2, $\beta_N$~1.35, $\beta_p$~1.55, lower signal null configuration (dRsep~-2cm), favorable Bt direction, heated by ~2.5MW ECRH at 140GHz and ~2MW LHCD at 4.6GHz. Good electron density feedback control is achieved through SMBI from the outer midplane. Divertor detachment control is achieved through feedback to the electron temperature measured by a divertor Langmuir probe at the outer target plate of lower divertor in this discharge. Detachment control through feedback to XUV radiation near the divertor X-point has also been achieved in other discharges. The electron temperature near the outer strike point of the lower divertor is maintained at ~2eV as measured by divertor Langmuir probes, indicating the achievement of divertor detachment. The divertor peak surface temperature measured by infrared camara is reduced from ~500 to ~250°C by the nitrogen seeding. Good energy confinement with $H_{98y2}$~1.10 is maintained for ~20s, then decreasing to $H_{98y2}$~1.05 until 50s when the discharge is terminated due to depletion of the volt-seconds. Long-pulse discharges of >70s (limited by volt-seconds) have been achieved with plasma current $I_p$ reduced to 350kA with $q_{95}$~7.0. However, the energy confinement is lower ($H_{98y2}$~1.0). Before the nitrogen seeding, grassy-ELM-like perturbations appear in the divertor D$\alpha$ signals as shown in Fig.1(d). Then, the perturbations are completely suppressed and a no-ELM regime is maintained till the end of the discharge. Tangsten and molybdenum impurity line emission saturate after 40s as well as the core radiation. Carbon impurity line emission keeps nearly constant. A broadband turbulence peaking at ~600kHz appears during the nitrogen seeding in the pedestal steep-gradient region near the pedestal top, which may be responsible for the ELM suppression. The improved particle confinement leads to a continuous decrease in the divertor D$\alpha$ signal, as well as a decrease in the SOL density, and more neutral particles are ionized in the pedestal region, resulting in an increase in the pedestal density gradient. This plasma regime has also been obtained at $I_p$=500kA and $q_{95}$~5.2 in 10s short pulses with $H_{98y2}$ up to 1.2. In summary, these experiments demonstrate a metal-wall compatible plasma regime for tokamak long-pulse operation with good energy confinement, stationary divertor detachment and no ELM, which are desired operational conditions for future fusion reactors.

Speaker: Guosheng Xu (Institute of Plasma Physics, Chinese Academy of Sciences)

11:25 How DIII-D Can Access New Plasma Regimes with More ECH to Close Long Pulse Fusion Pilot Plant Knowledge Gaps

The DIII-D research program will close key gaps in knowledge for the design and operation of long-pulse tokamak fusion pilot plants (FPPs) by deploying increased electron cyclotron heating and current drive power to reach and study more relevant plasma regimes. The U.S. is focused on low capital cost compact FPPs on the path to commercialization. Steady-state or long-pulse inductive tokamaks are a potential path, but these have four broad categories of knowledge gaps for processes at and inside the first wall: (1) Core-edge integration: how do we sustain high fusion power density in the core while exhausting heat and particles without damaging plasma-facing materials? (2) Core optimization: how do we achieve sufficient energy confinement and MHD stability for high fusion gain and high non-inductive current fractions for long-pulse or steady-state? (3) Detrimental transient avoidance: how do we avoid or suppress large NTMs and ELMs? (4) Plasma-material interactions: which first wall and divertor structural materials are acceptable? Adding a significant amount of flexible ECH power to DIII-D will enable access to new plasma target regimes designed to study and help close these gaps. Guided by publicly available compact FPP design studies, DIII-D plasma targets have been chosen that are designed to match a few but not all key FPP metrics simultaneously to study a particular gap question. For example, core-edge integration in an H-mode based, long-pulse, inductively sustained scenario would be well investigated with access to DIII-D plasmas having pedestal collisionality less than ~0.2, neutral penetration depth less than the pedestal width, betaT>=5%, H98y2>=1, and fG,ped>=0.9. This would allow assessment of edge plasma dynamics when the high-pressure pedestal structure is set by transport rather than deep neutral fueling, and would be an ideal regime to study radiative and/or detached divertor operation in. Other plasma targets are defined for investigating optimal core transport and stability with FPP relevant values, such as low rotation, Te/Ti>=1, and low fast ion fraction, and for investigating NTM suppression. Predictive integrated modelling using the physics-based IPS-FASTRAN code will be presented that quantifies – with systematic evaluation of uncertainties due to limited physics models - the minimum needs for additional ECH power for a few different target scenarios. This includes delivered power and current drive versus radius, gyrotron frequency, and X- or O-mode operation, assuming existing launch geometries. Preliminary modeling already predicts that with 20 MW of NBI power (another planned upgrade), 8.4 MW of 170 GHz X3-mode plus 5.6 MW of 137 GHz O2-mode would enable DIII-D plasmas close to or exceeding all the simultaneous metrics in the example above.

Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725

Speaker: Christopher Holcomb (Lawrence Livermore National Laboratory)

11:50 WEST operational domain predictions of L-mode non-inductive discharges with an integrated modeling approach

WEST is a machine designed for attaining Long Pulse Operation [1] with Lower-hybrid Heating and Current Drive (LHCD), superconducting coils, actively-cooled components and a W wall. In 2024, a more than six-minutes-long L-mode plasma was sustained with 1.15 GJ of injected energy. In parallel, developments were undertaken on the modeling side. The High Fidelity Pulse Simulator (HFPS) [2], an IMAS-coupled version of the integrated JINTRAC suite of codes, has been used to predict self-consistently plasma kinetic profiles and LHCD current drive efficiency for such long pulses. Predictions rely on the simultaneous evolution of the turbulent transport with TGLF-sat2 [3] and the LHCD source with METIS reduced model [4,5]. The latter estimates the current generated using a scaling law, derived from Tore Supra and JET databases analysis [6] and checked against WEST experimental data.

This workflow is used to accurately reproduce a reference experimental plasma. In particular, electron temperature and density profiles (Figure 1a and 1b), global energy confinement time and loop voltage are shown to be in good quantitative agreement. Then, the operational domain for a fully non-inductive discharge has been sought, scanning the LHCD injected power $P_{LHCD}$, the plasma current $I_P$ and the electron line-averaged density $n_{el}$. This exploratory modeling effort has been found in very good match with subsequent experimental ($P_{LHCD}, I_P$) scans (Figure 1c and 1d). Finally, using this integrated modeling framework, the impact of additional physics ingredients such as low-Z impurity seeding and central ECCD can be investigated prior to the next WEST experimental campaign.

Figure 1. Predicted (a) electron density profile against reflectometry measurements, and (b) electron temperature profile against ECE measurements for the reference discharge. Validation of LHCD power and plasma current scans against experiment for energy confinement time (c) and loop voltage (d).

[1]: X. Litaudon, et al., Nucl. Fusion 64, 015001 (2024).
[2]: See TSVV11 wikipages https://wiki.euro-fusion.org/wiki/TSVV-11.
[3]: C. Angioni, et al., Nucl. Fusion, 63, 126035 (2022).
[4]: J.-F. Artaud, et al., Nucl. Fusion 58, 105001 (2018).
[5]: R. Dumont, et al., Phys. Plasmas 7, 4972–4982 (2000).
[6]: M. Goniche, et al., AIP 787, 307–310 (2005).

Speaker: Theo Fonghetti (CEA Cadarache)

12:15 → 13:45 Lunch Break

13:45 → 14:35 RAMI & LPO session

13:45 Development of Long Pulse Fully Non-inductive High-confinement Plasma with Full Tungsten Limiter/Divertor on EAST

The major goal of EAST is to demonstrate long-pulse high-performance regime with tungsten wall for scientific understanding in support of future fusion device. Recently, one hundred-seconds steady-state plasmas adapted to reactor-relevant parameters have been achieved in high poloidal beta scenario: (1) a fully non-inductive plasma with high density (ne/nGW ~0.82) and high bootstrap current (fBS~56%) at high βp ~3.0; (2) excellent energy confinement quality (H98y2 ~1.5/βN ~1.8) by electron dominant heating with zero torque injection; (3) key issues of particle and heat balance tackled with actively cooling tungsten divertor; (4) small ELMs throughout the discharge compatible with detachment divertor sustained high core confinement. We enhanced H&CD efficiency at high density by applying high frequency low hybrid current drive (LHCD) system, together with low recycling wall and the synergy effect by adding electron cyclotron heating (ECH) system. The high energy confinement accompanying internal transport barrier is sustained by electron dominant heating, the broaden current profile with high-qmin and negative or weak shear was optimized by early heating and RF power deposition, the analysis also points to the strong effect of Shafranov shift on turbulence. In addition, key issues of particle and heat balance for long pulse operation are addressed with actively cooling tungsten divertor and real-time wall-conditioning. ELMs are kept at low amplitude level at high density to facilitate efficient RF power coupling and reduce divertor sputtering/erosion. Furthermore, long pulse H-modes up to 100s also have been successfully achieved under both boronized and uncoating wall conditions with type II ELMs at q95 ~ 6 similar to ITER SSO scheme. Energy confinement can be kept high (H98y2 >1.1) with and is independent of the tungsten source from the main limiter and heating mixed. On the contrary type-I ELMy H-modes in similar q95 ~ 6 conditions is hard to be maintained for high tungsten impurity sources, and decreased energy confinement illustrating the need for optimum ELM control with full tungsten wall. Solutions in these areas are vital and can offer unique contributions to the critical issues relevant for the next-step fusion such as ITER and CFETR.

Speaker: Mr Xianzu Gong (ASIPP)

14:10 Technical and Engineering challenges for long pulses on JET ITER Like Wall

The typical pulse on the JET tokamak is ~10s during the main phase of the discharge, however long discharge operation (>30s) is possible with sufficient preparation and care. During the last period of JET operation in 2023 long pulses in deuterium plasmas were developed to assess the sustainment of the plasma performance over several times the current resistive time scale and to address plasma-wall interaction physics in a full metallic environment with the ITER-like wall (ILW), with a W divertor and a Be first wall [1]. Two types of long duration discharges were successfully developed for this purpose: (i) a 30s ELMy H-mode with combined 12-14MW neutral beam heating (NBI) and 2MW of ion-cyclotron resonance heating (ICRH) and (ii) a 60s long pulse with 4-5MW of NBI and 2MW of ICRH. Both operational scenarios are based on previously developed hybrid-like plasmas at JET [2] with IP=1.4MA and B0=1.9T (q954). This was the first time that NBI could be used for durations longer than 30s on JET.
To prepare for the long pulse operation an analysis of heatloads was required to ensure the pulse was safe for the machine, this defined a number of choices on Bt and plasma configuration. While the 30s pulse was within the control and protection systems commissioned operating envelope the 60s pulse was beyond these systems normal operation. The control, diagnostic and protection systems were adapted and tested as far as possible to ensure they would work in the real pulse and a number of issues resolved over a series of ‘dry runs’. A broad team of experts was convened to review all these aspects and support the pulses on the days they were carried out. Significant modifications were required to carry out the experiment which had to be reversed before going back to standard operations. Even with these extensive preparations issues were found in the pulses, in particular on the heating systems and plasma shape controller. These were resolved leading to the success of the 60s pulse.
The technical details of these preparations and their implementation will be presented in this contribution while the results of the pulses will be shown separately [3].

[1] G. F. Matthews et al 2011 Phys. Scr. 2011 014001
[2] J. Hobirk et al 2023 Nucl. Fusion 63 112001
[3] E. Lerche – Long pulse operation with the JET ITER-Like Wall this meeting

Speaker: Dr damian king (UKAEA)

14:35 → 15:05 Coffee Break

15:05 → 16:35 Discussion on critical issues for Long Pulse Operation focusing on issues reported during sessions on topics LPO, PWI, Control, H&CD, RAMI & N and closing

Summary-Discussion_2022-vf.pdf

Friday, 18 October

09:30 → 12:00 CICLOP session