According to recent DIII-D experiments (Logan et al 2024 Nucl. Fusion64 014003), injecting edge localized electron cyclotron current drive (ECCD) in the counter-plasma-current (counter-Ip) direction reduces the n = 3 resonant magnetic perturbation (RMP) current threshold for edge-localized mode (ELM) suppression, while co-Ip ECCD during the suppressed ELM phase causes a back transition to ELMing. This paper presents nonlinear two-fluid simulations on the ECCD manipulation of edge magnetic islands induced by RMP using the TM1 code. In the presence of a magnetic island chain at the pedestal-top, co-Ip ECCD is found to decrease the island width and restore the initially degraded pedestal pressure when its radial deposition location is close to the rational surface of the island. With a sufficiently strong co-Ip ECCD current, the RMP-driven magnetic island can be healed, and the pedestal pressure fully recovers to its initial ELMing state. On the contrary, counter-Ip ECCD is found to increase the island width and further reduce the pedestal pressure to levels significantly below the peeling-ballooning-mode limited height, leading to even stationary ELM suppression. These simulations align with the results from DIII-D experiments. However, when multiple magnetic island chains are present at the pedestal-top, the ECCD current experiences substantial broadening, and its effects on the island width and pedestal pressure become negligible. Further simulations reveal that counter-Ip ECCD enhances RMP penetration by lowering the penetration threshold, with the degree of reduction proportional to the amplitude of ECCD current. For the ∼1 MW ECCD in DIII-D, the predicted decrease in the RMP penetration threshold for ELM suppression is approximately 20%, consistent with experimental observations. These simulations indicate that edge-localized ECCD can be used to either facilitate RMP-driven ELM suppression or optimize the confinement degradation.
ISSN: 1741-4326
Nuclear Fusion is the acknowledged world-leading journal specializing in fusion. The journal covers all aspects of research, theoretical and practical, relevant to controlled thermonuclear fusion.
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Q.M. Hu et al 2024 Nucl. Fusion 64 046027
T. Qian et al 2022 Nucl. Fusion 62 084001
A first-of-a-kind optimized stellarator for confining plasma has been designed and is being constructed with planar circular coils and permanent magnets composed of identical elements. The equilibrium is optimized to be quasi-axisymmetric for good particle confinement. The combination of permanent magnets and planar coils is significantly simpler to construct than fabricating three-dimensionally shaped coils, yet they are able to produce lower helical magnetic ripple than existing devices by two orders of magnitude in , a characteristic neoclassical transport metric.
Sehila M. Gonzalez de Vicente et al 2022 Nucl. Fusion 62 085001
In the absence of official standards and guidelines for nuclear fusion plants, fusion designers adopted, as far as possible, well-established standards for fission-based nuclear power plants (NPPs). This often implies interpretation and/or extrapolation, due to differences in structures, systems and components, materials, safety mitigation systems, risks, etc. This approach could result in the consideration of overconservative measures that might lead to an increase in cost and complexity with limited or negligible improvements. One important topic is the generation of radioactive waste in fusion power plants. Fusion waste is significantly different to fission NPP waste, i.e. the quantity of fusion waste is much larger. However, it mostly comprises low-level waste (LLW) and intermediate level waste (ILW). Notably, the waste does not contain many long-lived isotopes, mainly tritium and other activation isotopes but no-transuranic elements. An important benefit of fusion employing reduced-activation materials is the lower decay heat removal and rapid radioactivity decay overall. The dominant fusion wastes are primarily composed of structural materials, such as different types of steel, including reduced activation ferritic martensitic steels, such as EUROFER97 and F82H, AISI 316L, bainitic, and JK2LB. The relevant long-lived radioisotopes come from alloying elements, such as niobium, molybdenum, nickel, carbon, nitrogen, copper and aluminum and also from uncontrolled impurities (of the same elements, but also, e.g. of potassium and cobalt). After irradiation, these isotopes might preclude disposal in LLW repositories. Fusion power should be able to avoid creating high-level waste, while the volume of fusion ILW and LLW will be significant, both in terms of pure volume and volume per unit of electricity produced. Thus, efforts to recycle and clear are essential to support fusion deployment, reclaim resources (through less ore mining) and minimize the radwaste burden for future generations.
Vignesh Gopakumar et al 2024 Nucl. Fusion 64 056025
Predicting plasma evolution within a Tokamak reactor is crucial to realizing the goal of sustainable fusion. Capabilities in forecasting the spatio-temporal evolution of plasma rapidly and accurately allow us to quickly iterate over design and control strategies on current Tokamak devices and future reactors. Modelling plasma evolution using numerical solvers is often expensive, consuming many hours on supercomputers, and hence, we need alternative inexpensive surrogate models. We demonstrate accurate predictions of plasma evolution both in simulation and experimental domains using deep learning-based surrogate modelling tools, viz., Fourier neural operators (FNO). We show that FNO has a speedup of six orders of magnitude over traditional solvers in predicting the plasma dynamics simulated from magnetohydrodynamic models, while maintaining a high accuracy (Mean Squared Error in the normalised domain ). Our modified version of the FNO is capable of solving multi-variable Partial Differential Equations, and can capture the dependence among the different variables in a single model. FNOs can also predict plasma evolution on real-world experimental data observed by the cameras positioned within the MAST Tokamak, i.e. cameras looking across the central solenoid and the divertor in the Tokamak. We show that FNOs are able to accurately forecast the evolution of plasma and have the potential to be deployed for real-time monitoring. We also illustrate their capability in forecasting the plasma shape, the locations of interactions of the plasma with the central solenoid and the divertor for the full (available) duration of the plasma shot within MAST. The FNO offers a viable alternative for surrogate modelling as it is quick to train and infer, and requires fewer data points, while being able to do zero-shot super-resolution and getting high-fidelity solutions.
M.S. Islam et al 2024 Nucl. Fusion 64 056036
The SOLPS-ITER code is utilized to analyze the boundary plasma associated with a fast-flow lithium (Li) divertor configuration in the fusion nuclear science facility (FNSF) tokamak and identify operational regimes with acceptable divertor and core conditions. Plasma transport from the SOLPS-ITER code has been coupled with a liquid metal (LM) MHD/heat transfer code to model a Li open-surface divertor design and assess its impact on the scrape-off-layer (SOL) and core plasma performance. Simulations with only Neon (Ne) impurity seeding have been conducted to evaluate its impact on meeting FNSF design demands for the divertor and upstream plasma parameters. Simulation results indicate that Ne seeding significantly mitigates divertor heat flux but potentially reduces both upstream electron and main ion density due to fuel dilution. The combined application of Ne seeding and deuterium (D2) puffing is required to satisfy the FNSF design requirements on upstream density ( ∼1× 1020 m−3) and divertor energy flux (10 MW m−2). D2 puffing plays a role in counteracting upstream density drops and augmenting energy and momentum losses through atomic and molecular processes.
The inlet Li flow velocity is systematically varied across a wide range to identify acceptable flows and corresponding LM surface temperatures. This comprehensive analysis identifies the acceptable Li flow parameters, LM surface temperature, and emitted Li fluxes necessary to meet the major design constraints. The emitted Li fluxes exhibit minimal impact on the main plasma at surface temperatures up to approximately ∼525 ∘C, corresponding emitted Li fluxes of up to φLi ∼2 atoms s−1. Uncertainties in the Li emission processes from the surface are also investigated, primarily influencing Li loss in the lower surface temperature range (C), with simulation results indicating a minor impact on the divertor and upstream plasma. Conversely, evaporation predominantly drives the Li loss processes at higher surface temperature ranges (C), contaminating both the divertor and upstream plasma.
G. Federici et al 2024 Nucl. Fusion 64 036025
High temperature superconductors (HTSs) offer the promise of operating at higher magnetic field and temperature. Recently, the use of high field magnets (by adopting HTS) has been promoted by several groups around the world, including new start-up entries, both to substantially reduce the size of a fusion power reactor system and as a breakthrough innovation that could dramatically accelerate fusion power deployment. This paper describes the results of an assessment to understand the impact of using high field magnets in the design of DEMO in Europe, considering a comprehensive list of physics and engineering limitations together with the interdependencies with other important parameters. Based on the results, it is concluded that increasing the magnetic field does not lead to a reduction in device size with relevant nuclear performance requirements, because (i) large structures are needed to withstand the enormous electromagnetic forces, (ii) thick blanket and n-shield structures are needed to protect the coils from radiation damage effects, and (iii) new divertor solutions with performances well beyond today's concepts are needed. Stronger structural materials allow for more compact tokamaks, but do not change the conclusion that scalability is not favourable when increasing the magnetic field, beyond a certain point, the machine size cannot be further reduced. More advanced structural support concepts for high-field coils have been explored and concluded that these solutions are either unfeasible or provide only marginal size reduction, by far not sufficient to account for the potential of operating at very high field provided by HTS. Additionally, the cost of high field coils is significant at today's price levels and shows to scale roughly with the square of the field. Nevertheless, it is believed that even when not operated at high field and starting within conventional insulated coils, HTS can still offer certain benefits. These include the simplification of the magnet cooling scheme thanks to increased temperature margin (indirect conduction cooling). This in turn can greatly simplify coil construction and minimize high-voltage risks at the terminals.
Santanu Banerjee et al 2024 Nucl. Fusion 64 046026
We present observations, numerical simulations, and analysis from experiments in the Lithium Tokamak Experiment-Beta (LTX-β) in which the electron temperature profile (Te(r)) shifts from flat to peaked and a tearing mode is also destabilized when the average density (neave) exceeds ∼1019 m−3. Flat Te(r) is obtained routinely in LTX-β, with a lithium coated, low-recycling first wall, once the external fueling is stopped and density decays [Boyle et al 2023 Nucl. Fusion 63 056020]. In the present experiment, flat Te profiles can be sustained while maintaining constant neave below a line averaged density threshold (neaveth) of ∼1019 m−3. Above neaveth, Te(r) shifts from flat to peaked and a tearing mode is destabilized. Due to low recycling, the achieved neave can be controlled precisely by external fueling and hence, a certain threshold of the edge neutral inventory from the external fueling is experimentally manifested through neaveth. The goal of the present work is to investigate the role of edge neutrals in determining Te(r) and MHD stability in the unique low-recycling regime of LTX-β. Our hypothesis is that the peaking of Te(r) beyond neaveth is due ultimately to the edge cooling by the cold neutrals beyond a critical fueling flux. At lower fueling flux, flat Te(r) results in broader pressure profile and lower resistivity, which in turn stabilizes the tearing mode. This hypothesis is supported by edge neutral density estimation by DEGAS 2 code. Mode analysis by singular value decomposition confirms the tearing mode structure to be m/n = 2/1 (m and n being the poloidal and toroidal mode numbers). Linear tearing stability analysis with M3D-C1 predicts that plasmas with neave> 1019 are highly susceptible to a n = 1 tearing mode. ORBIT simulations, however, confirmed that the tearing modes do not contribute to the loss of fast ions from neutral beam injection. This study shows for the first time that the neutral inventory at the edge could be one of the deciding factors for the achievability of the unique operation regime of flat Te(r) and the excitation of tearing activity that could be disruptive for the plasmas.
J. Mailloux et al 2022 Nucl. Fusion 62 042026
The JET 2019–2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major neutral beam injection upgrade providing record power in 2019–2020, and tested the technical and procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle (α) physics in the coming D–T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed shattered pellet injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design and operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D–T benefited from the highest D–D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.
M. Hoelzl et al 2021 Nucl. Fusion 61 065001
JOREK is a massively parallel fully implicit non-linear extended magneto-hydrodynamic (MHD) code for realistic tokamak X-point plasmas. It has become a widely used versatile simulation code for studying large-scale plasma instabilities and their control and is continuously developed in an international community with strong involvements in the European fusion research programme and ITER organization. This article gives a comprehensive overview of the physics models implemented, numerical methods applied for solving the equations and physics studies performed with the code. A dedicated section highlights some of the verification work done for the code. A hierarchy of different physics models is available including a free boundary and resistive wall extension and hybrid kinetic-fluid models. The code allows for flux-surface aligned iso-parametric finite element grids in single and double X-point plasmas which can be extended to the true physical walls and uses a robust fully implicit time stepping. Particular focus is laid on plasma edge and scrape-off layer (SOL) physics as well as disruption related phenomena. Among the key results obtained with JOREK regarding plasma edge and SOL, are deep insights into the dynamics of edge localized modes (ELMs), ELM cycles, and ELM control by resonant magnetic perturbations, pellet injection, as well as by vertical magnetic kicks. Also ELM free regimes, detachment physics, the generation and transport of impurities during an ELM, and electrostatic turbulence in the pedestal region are investigated. Regarding disruptions, the focus is on the dynamics of the thermal quench (TQ) and current quench triggered by massive gas injection and shattered pellet injection, runaway electron (RE) dynamics as well as the RE interaction with MHD modes, and vertical displacement events. Also the seeding and suppression of tearing modes (TMs), the dynamics of naturally occurring TQs triggered by locked modes, and radiative collapses are being studied.
P. Rodriguez-Fernandez et al 2022 Nucl. Fusion 62 042003
The SPARC tokamak project, currently in engineering design, aims to achieve breakeven and burning plasma conditions in a compact device, thanks to new developments in high-temperature superconductor technology. With a magnetic field of 12.2 T on axis and 8.7 MA of plasma current, SPARC is predicted to produce 140 MW of fusion power with a plasma gain of Q ≈ 11, providing ample margin with respect to its mission of Q > 2. All tokamak systems are being designed to produce this landmark plasma discharge, thus enabling the study of burning plasma physics and tokamak operations in reactor relevant conditions to pave the way for the design and construction of a compact, high-field fusion power plant. Construction of SPARC is planned to begin by mid-2021.
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M.I. Kobayashi et al 2024 Nucl. Fusion 64 066026
This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.
G. Wang et al 2024 Nucl. Fusion 64 066024
Sawteeth are one of the concerning instabilities in ITER and future burning plasma experiments. Sawtooth dynamics and its interaction with broadband plasma turbulence has been a challenge for predictive simulations of core transport in future fusion devices. This study provides new observations of core turbulence behavior during sawtooth oscillations in DIII-D hydrogen L-mode neutral beam injection heated plasmas in an inner wall limited configuration. A strong correlation of electron temperature and density turbulence levels with the sawtooth oscillation phase has been observed at locations inside the Te inversion radius and/or safety factor q = 1 magnetic surface. The Te turbulence amplitude in the core during the sawtooth ramp exhibits a critical Te gradient behavior inside but not near the Te inversion radius/q = 1 magnetic surface. The most unstable mode calculated from the trapped gyro-landau fluid turbulence simulations reveal a change from low-k ion-type to low-k electron-type modes from pre- to post- sawtooth crash time periods.
L.N. Liu et al 2024 Nucl. Fusion 64 066025
Ion Cyclotron Resonance Heating (ICRH) has been a dependable tool for sturdy plasma heating with high RF power of several megawatts. However, a sudden increase in the reflected power during ICRH heating experiments is a problem that should be solved for future fusion experimental devices. To solve this issue, a load tolerant matching network has been designed for the ICRH system in EAST. The matching network includes a 3-stub tuner impedance matching system with conjugate-T structure, 30 Ω to 50 Ω transmission line and center grounded antenna strap. By maintaining a low reflection ratio in the network for a wide range of resistance, this matching network can allow sturdy high-power operations without fast impedance matching in EAST. In our matching network, the two arms of a conjugate-T are designed to have a length difference which could mitigate current imbalance and antenna poloidal phasing outside of the control problem. The T-point corresponds to the maximum point of the standing wave voltage, which could greatly improve the input impedance of the antenna.
Masakatsu Fukumoto et al 2024 Nucl. Fusion 64 066022
Electron cyclotron wall conditioning with neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with O-mode EC waves including X-mode polarization with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.
F. Kin et al 2024 Nucl. Fusion 64 066023
The avalanche type of transport can induce a long-radial transport and thus can contribute to the global profile formation. In this study, we observed the heat perturbations exhibiting avalanche-like transport in the stellarator/heliotron device, Heliotron J, and the tokamak device, JT-60U. We found that the electron heat propagation in Heliotron J is mainly generated from the heating source region. The relatively high value of the Hurst exponent, which is a signature of avalanches, depends on the total heating power. On the other hand, the electron and ion heat avalanches measured in JT-60U tend to spread from the local peak of the temperature gradient and are not influenced by the heating source profiles. The contrasting features of avalanches in stellarator/heliotrons and tokamaks potentially imply the difference in the temperature profile formation, such as the presence of stiffness.
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G.D. Conway et al 2022 Nucl. Fusion 62 013001
Geodesic acoustic modes (GAMs) are ubiquitous oscillatory flow phenomena observed in toroidal magnetic confinement fusion plasmas, such as tokamaks and stellarators. They are recognized as the non-stationary branch of the turbulence driven zonal flows which play a critical regulatory role in cross-field turbulent transport. GAMs are supported by the plasma compressibility due to magnetic geodesic curvature—an intrinsic feature of any toroidal confinement device. GAMs impact the plasma confinement via velocity shearing of turbulent eddies, modulation of transport, and by providing additional routes for energy dissipation. GAMs can also be driven by energetic particles (so-called EGAMs) or even pumped by a variety of other mechanisms, both internal and external to the plasma, opening-up possibilities for plasma diagnosis and turbulence control. In recent years there have been major advances in all areas of GAM research: measurements, theory, and numerical simulations. This review assesses the status of these developments and the progress made towards a unified understanding of the GAM behaviour and its role in plasma confinement. The review begins with tutorial-like reviews of the basic concepts and theory, followed by a series of topic orientated sections covering different aspects of the GAM. The approach adopted here is to present and contrast experimental observations alongside the predictions from theory and numerical simulations. The review concludes with a comprehensive summary of the field, highlighting outstanding issues and prospects for future developments.
L. Marrelli et al 2021 Nucl. Fusion 61 023001
This paper reviews the research on the reversed field pinch (RFP) in the last three decades. Substantial experimental and theoretical progress and transformational changes have been achieved since the last review (Bodin 1990 Nucl. Fusion 30 1717–37). The experiments have been performed in devices with different sizes and capabilities. The largest are RFX-mod in Padova (Italy) and MST in Madison (USA). The experimental community includes also EXTRAP-T2R in Sweden, RELAX in Japan and KTX in China. Impressive improvements in the performance are the result of exploration of two lines: the high current operation (up to 2 MA) with the spontaneous occurrence of helical equilibria with good magnetic flux surfaces and the active control of the current profile. A crucial ingredient for the advancements obtained in the experiments has been the development of state-of-art active feedback control systems allowing the control of MHD instabilities in presence of a thin shell. The balance between achievements and still open issues leads us to the conclusion that the RFP can be a valuable and diverse contributor in the quest for fusion electricity.
Mohamed Abdou et al 2021 Nucl. Fusion 61 013001
The tritium aspects of the DT fuel cycle embody some of the most challenging feasibility and attractiveness issues in the development of fusion systems. The review and analyses in this paper provide important information to understand and quantify these challenges and to define the phase space of plasma physics and fusion technology parameters and features that must guide a serious R&D in the world fusion program. We focus in particular on components, issues and R&D necessary to satisfy three 'principal requirements': (1) achieving tritium self-sufficiency within the fusion system, (2) providing a tritium inventory for the initial start-up of a fusion facility, and (3) managing the safety and biological hazards of tritium. A primary conclusion is that the physics and technology state-of-the-art will not enable DEMO and future power plants to satisfy these principal requirements. We quantify goals and define specific areas and ideas for physics and technology R&D to meet these requirements. A powerful fuel cycle dynamics model was developed to calculate time-dependent tritium inventories and flow rates in all parts and components of the fuel cycle for different ranges of parameters and physics and technology conditions. Dynamics modeling analyses show that the key parameters affecting tritium inventories, tritium start-up inventory, and tritium self-sufficiency are the tritium burn fraction in the plasma (fb), fueling efficiency (ηf), processing time of plasma exhaust in the inner fuel cycle (tp), reactor availability factor (AF), reserve time (tr) which determines the reserve tritium inventory needed in the storage system in order to keep the plant operational for time tr in case of any malfunction of any part of the tritium processing system, and the doubling time (td). Results show that ηffb > 2% and processing time of 1–4 h are required to achieve tritium self-sufficiency with reasonable confidence. For ηffb = 2% and processing time of 4 h, the tritium start-up inventory required for a 3 GW fusion reactor is ∼11 kg, while it is <5 kg if ηffb = 5% and the processing time is 1 h. To achieve these stringent requirements, a serious R&D program in physics and technology is necessary. The EU-DEMO direct internal recycling concept that carries fuel directly from the plasma exhaust gas to the fueling systems without going through the isotope separation system reduces the overall processing time and tritium inventories and has positive effects on the required tritium breeding ratio (TBRR). A significant finding is the strong dependence of tritium self-sufficiency on the reactor availability factor. Simulations show that tritium self-sufficiency is: impossible if AF < 10% for any ηffb, possible if AF > 30% and 1% ⩽ ηffb ⩽ 2%, and achievable with reasonable confidence if AF > 50% and ηffb > 2%. These results are of particular concern in light of the low availability factor predicted for the near-term plasma-based experimental facilities (e.g. FNSF, VNS, CTF), and can have repercussions on tritium economy in DEMO reactors as well, unless significant advancements in RAMI are made. There is a linear dependency between the tritium start-up inventory and the fusion power. The required tritium start-up inventory for a fusion facility of 100 MW fusion power is as small as 1 kg. Since fusion power plants will have large powers for better economics, it is important to maintain a 'reserve' tritium inventory in the tritium storage system to continue to fuel the plasma and avoid plant shutdown in case of malfunctions of some parts of the tritium processing lines. But our results show that a reserve time as short as 24 h leads to unacceptable reserve and start-up inventory requirements. Therefore, high reliability and fast maintainability of all components in the fuel cycle are necessary in order to avoid the need for storing reserve tritium inventory sufficient for continued fusion facility operation for more than a few hours. The physics aspects of plasma fueling, tritium burn fraction, and particle and power exhaust are highly interrelated and complex, and predictions for DEMO and power reactors are highly uncertain because of lack of experiments with burning plasma. Fueling by pellet injection on the high field side of tokamak has evolved to be the preferred method to fuel a burning plasma. Extrapolation from the DIII-D penetration scaling shows fueling efficiency expected in DEMO to be <25%, but such extrapolations are highly uncertain. The fueling efficiency of gas in a reactor relevant regime is expected to be extremely poor and not very useful for getting tritium into the core plasma efficiently. Gas fueling will nonetheless be useful for feedback control of the divertor operating parameters. Extensive modeling has been carried out to predict burn fraction, fueling requirements, and fueling efficiency for ITER, DEMO, and beyond. The fueling rate required to operate Q = 10 ITER plasmas in order to provide the required core fueling, helium exhaust and radiative divertor plasma conditions for acceptable divertor power loads was calculated. If this fueling is performed with a 50–50 DT mix, the tritium burn fraction in ITER would be ∼0.36%, which is too low to satisfy the self-sufficiency conditions derived from the dynamics modeling for fusion reactors. Extrapolation to DEMO using this approach would also yield similarly low burn fraction. Extensive analysis presented shows that specific features of edge neutral dynamics in ITER and fusion reactors, which are different from present experiments, open possibilities for optimization of tritium fueling and thus to improve the burn fraction. Using only tritium in pellet fueling of the plasma core, and only deuterium for edge density, divertor power load and ELM control results in significant increase of the burn fraction to 1.8–3.6%. These estimates are performed with physics models whose results cannot be fully validated for ITER and DEMO plasma conditions since these cannot be achieved in present tokamak experiments. Thus, several uncertainties remain regarding particle transport and scenario requirements in ITER and DEMO. The safety standard requirements for protection of the public and release guidelines for tritium have been reviewed. General safety approaches including minimizing tritium inventories, reducing tritium permeation through materials, and decontaminating material for waste disposal have been suggested.
Boris N. Breizman et al 2019 Nucl. Fusion 59 083001
Of all electrons, runaway electrons have long been recognized in the fusion community as a distinctive population. They now attract special attention as a part of ITER mission considerations. This review covers basic physics ingredients of the runaway phenomenon and the ongoing efforts (experimental and theoretical) aimed at runaway electron (RE) taming in the next generation tokamaks. We emphasize the prevailing physics themes of the last 20 years: the hot-tail mechanism of runaway production, RE interaction with impurity ions, the role of synchrotron radiation in runaway kinetics, RE transport in presence of magnetic fluctuations, micro-instabilities driven by REs in magnetized plasmas, and vertical stability of the plasma with REs. The review also discusses implications of the runaway phenomenon for ITER and the current strategy of RE mitigation.
M.K.A. Thumm et al 2019 Nucl. Fusion 59 073001
In many tokamak and stellarator experiments around the globe that are investigating energy production via controlled thermonuclear fusion, electron cyclotron heating and current drive (ECH&CD) are used for plasma start-up, heating, non-inductive current drive and magnetohydrodynamic stability control. ECH will be the first auxiliary heating method used on ITER. Megawatt-class, continuous wave gyrotrons are employed as high-power millimeter (mm)-wave sources. The present review reports on the worldwide state-of-the-art of high-power gyrotrons. Their successful development during recent years changed ECH from a minor to a major heating method. After a general introduction of the various functions of ECH&CD in fusion physics, especially for ITER, section 2 will explain the fast-wave gyrotron interaction principle. Section 3 discusses innovations on the components of modern long-pulse fusion gyrotrons (magnetron injection electron gun, beam tunnel, cavity, quasi-optical output coupler, synthetic diamond output window, single-stage depressed collector) and auxiliary components (superconducting magnets, gyrotron diagnostics, high-power calorimetric dummy loads). Section 4 deals with present megawatt-class gyrotrons for ITER, W7-X, LHD, EAST, KSTAR and JT-60SA, and also includes tubes for moderate pulse length machines such as ASDEX-U, DIII-D, HL-2A, TCV, QUEST and GAMMA-10. In section 5 the development of future advanced fusion gyrotrons is discussed. These are tubes with higher frequencies for DEMO, multi-frequency (multi-purpose) gyrotrons, stepwise frequency tunable tubes for plasma stabilization, injection-locked and coaxial-cavity multi-megawatt gyrotrons, as well as sub-THz gyrotrons for collective Thomson scattering. Efficiency enhancement via multi-stage depressed collectors, fast oscillation recovery methods and reliability, availability, maintainability and inspectability will be discussed at the end of this section.
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Yin et al
Efficient ion heating is crucial for future fusion devices, and the only way to heat ions directly is ion cyclotron resonance heating (ICRH). Reported here is a full wave solver integrated with a Fokker–Planck code for optimizing ion heating with ion cyclotron range of frequency waves for the International Thermonuclear Experimental Reactor (ITER)deuterium-tritium plasma. Both the direct absorption of minority ions and the power transfer to bulk ions via collisions are considered, while also accounting for the edge effects on ion absorption near the core. The simulation results show that the appropriate scrape-off layer density profile and parallel wave number lead to enhanced edge coupling and broaden the absorption region with moderate absorption intensity of the minority ions, which is very important for ion heating. More power from the heated ions is transferred to bulk ions than to electrons through collisions in our simulation via optimization, and reducing the total RF power results in a significant increase of the absorbed fraction of bulk ions.
Terranova et al
The RFX-mod2 installation is planned to be completed by 2024 and the start of operations is expected in 2025. The high flexibility of the machine (already tested in the previous RFX-mod experiment) allows operation in Reversed Field Pinch and tokamak configuration as well as ultra-low q pulses. In this work we present predictive analysis on transport, performances and plasma control in RFX-mod2 in view of the first experimental campaigns.
Ding et al
In order to sustain good LHW (lower hybrid wave) -plasma coupling for long pulse plasma operation, 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 such feedback control can work 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. Experiments between SMBI puffing and the gas-puffing fed by piezoelectric valve near the antenna are further investigated, showing that the response time on RC 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 such feedback control is effective for long pulse LHW-plasma coupling and the gas puffing by SMBI in the electric drift side of LHW antenna could offers an effective way to sustain good LHW coupling in steady state operation in future. Further optimization will be continued later.
Zhang et al
The stabilization of the m/n=2/1 neoclassical tearing mode (NTM) by electron cyclotron current drive (ECCD) has been carried out in EAST H-mode discharges, where m/n is the poloidal/toroidal mode number. The major experimental results are reported for the first time in this paper. To facilitate the experimental study, the magnetic island (NTM) is generated by a sufficiently large amplitude of the externally applied resonant magnetic perturbation (RMP). After switching off the RMP, the NTM exists due to the bootstrap current perturbation, with the magnetic island width being about 5 cm for the local equilibrium bootstrap current fraction being larger than 10 %. By applying the localized ECCD later, the NTM is fully suppressed if the radial misalignment between the magnetic island and the ECCD location is sufficiently small. The stabilizing effect depends on both the radial misalignment and the applied electron cyclotron wave power. More importantly, it is found that the NTM can be avoided when applying ECCD earlier during the ramp-up phase of the RMP amplitude, if ECCD is localized around the O-point of the magnetic island, indicating an efficient way for avoiding locked modes that can lead to the major disruptions of tokamak plasmas.
Schissel et al
Full remote scientific operation of the DIII-D National Fusion Facility is now possible through significant advances in the computer science hardware and software infrastructure made over the last decade. Capabilities around information visualization, data movement, and communication have all been enhanced. The level of capability deployed to remotely operate DIII-D required an infrastructure advancement over what had previously been achieved in the fusion community. The large quantity of real-time data that is automatically displayed on DIII-D's control room screens can now be visualized by remote participants via web-based applications. New audio/video solutions using the VoIP and instant messaging application Discord have been implemented to mimic the dynamic and ad hoc scientific conversations that are critical in successfully operating an experimental campaign. Discord's ability for a user to rapidly move between audio channels, text with images, and share screens is a significant enhancement over traditional videoconferencing tools. In addition, multiple combinations of broadcast audio are made available via a web-based application to allow remote participants to simultaneously listen to general announcements/sounds while conducting their own specific conversations. Secure methodologies have been put into place to allow remote control of hardware including DIII-D's plasma control system application. Secure methods also included the ability of the on-site team to closely coordinate their work with remote team members which has been enhanced through extensions to the wireless network and the use of tablet computers for audio/video/screen sharing. However, no amount of software can fully replace the need for "hands on hardware." This infrastructure was severely stress tested during the COVID-19 pandemic where occupancy of the DIII-D control room was restricted. Operational efficiency during the pandemic, measured in discharges per hour, remained high (3.8 +/- 0.8) compared to values obtained pre-pandemic (3.7 +/- 0.8).
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M.I. Kobayashi et al 2024 Nucl. Fusion 64 066026
This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.
G. Wang et al 2024 Nucl. Fusion 64 066024
Sawteeth are one of the concerning instabilities in ITER and future burning plasma experiments. Sawtooth dynamics and its interaction with broadband plasma turbulence has been a challenge for predictive simulations of core transport in future fusion devices. This study provides new observations of core turbulence behavior during sawtooth oscillations in DIII-D hydrogen L-mode neutral beam injection heated plasmas in an inner wall limited configuration. A strong correlation of electron temperature and density turbulence levels with the sawtooth oscillation phase has been observed at locations inside the Te inversion radius and/or safety factor q = 1 magnetic surface. The Te turbulence amplitude in the core during the sawtooth ramp exhibits a critical Te gradient behavior inside but not near the Te inversion radius/q = 1 magnetic surface. The most unstable mode calculated from the trapped gyro-landau fluid turbulence simulations reveal a change from low-k ion-type to low-k electron-type modes from pre- to post- sawtooth crash time periods.
L.N. Liu et al 2024 Nucl. Fusion 64 066025
Ion Cyclotron Resonance Heating (ICRH) has been a dependable tool for sturdy plasma heating with high RF power of several megawatts. However, a sudden increase in the reflected power during ICRH heating experiments is a problem that should be solved for future fusion experimental devices. To solve this issue, a load tolerant matching network has been designed for the ICRH system in EAST. The matching network includes a 3-stub tuner impedance matching system with conjugate-T structure, 30 Ω to 50 Ω transmission line and center grounded antenna strap. By maintaining a low reflection ratio in the network for a wide range of resistance, this matching network can allow sturdy high-power operations without fast impedance matching in EAST. In our matching network, the two arms of a conjugate-T are designed to have a length difference which could mitigate current imbalance and antenna poloidal phasing outside of the control problem. The T-point corresponds to the maximum point of the standing wave voltage, which could greatly improve the input impedance of the antenna.
Masakatsu Fukumoto et al 2024 Nucl. Fusion 64 066022
Electron cyclotron wall conditioning with neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with O-mode EC waves including X-mode polarization with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.
F. Kin et al 2024 Nucl. Fusion 64 066023
The avalanche type of transport can induce a long-radial transport and thus can contribute to the global profile formation. In this study, we observed the heat perturbations exhibiting avalanche-like transport in the stellarator/heliotron device, Heliotron J, and the tokamak device, JT-60U. We found that the electron heat propagation in Heliotron J is mainly generated from the heating source region. The relatively high value of the Hurst exponent, which is a signature of avalanches, depends on the total heating power. On the other hand, the electron and ion heat avalanches measured in JT-60U tend to spread from the local peak of the temperature gradient and are not influenced by the heating source profiles. The contrasting features of avalanches in stellarator/heliotrons and tokamaks potentially imply the difference in the temperature profile formation, such as the presence of stiffness.
Cong Li et al 2024 Nucl. Fusion 64 068001
This comment examines a recent study claiming the observation of deuterium (D) supersaturated surface layer (SSL) in tungsten samples exposed to D plasma in the PISCES-A plasma device at ion energies as low as 45 eV/D (Nishijima et al 2023 Nucl. Fusion63 126003). Applying SDTrimSP simulations and recalling the previous study on SSL formation, herein we want to emphasize that the proposed model in the mentioned paper relies on a number of strong assumptions and that many of the observations made can be more easily rationalized by the presence of impurities. The present comment will be conducive to avoid some possible misunderstanding on the SSL formation mechanism.
D. Nishijima et al 2024 Nucl. Fusion 64 068002
We reply to the comment by Li et al (submitted to Nucl. Fusion with this response) on our recent paper Nishijima et al (2023 Nucl. Fusion63 126003). In this response, we address the existence of an incident ion energy, Ei, threshold for the deuterium (D) supersaturated surface layer (DSSL) formation with a newly conducted D plasma exposure experiment at Ei ∼ 20 eV. It is also further demonstrated, based on new experiments where the ion flux, fluence, and sample temperature are scanned, that non-kinetic (ballistic) processes play a role in the DSSL formation and growth. In addition, the effect of impurities in our plasma is discussed also with a new analysis of the surface composition made after a D plasma exposure.
Lan Yin et al 2024 Nucl. Fusion
Efficient ion heating is crucial for future fusion devices, and the only way to heat ions directly is ion cyclotron resonance heating (ICRH). Reported here is a full wave solver integrated with a Fokker–Planck code for optimizing ion heating with ion cyclotron range of frequency waves for the International Thermonuclear Experimental Reactor (ITER)deuterium-tritium plasma. Both the direct absorption of minority ions and the power transfer to bulk ions via collisions are considered, while also accounting for the edge effects on ion absorption near the core. The simulation results show that the appropriate scrape-off layer density profile and parallel wave number lead to enhanced edge coupling and broaden the absorption region with moderate absorption intensity of the minority ions, which is very important for ion heating. More power from the heated ions is transferred to bulk ions than to electrons through collisions in our simulation via optimization, and reducing the total RF power results in a significant increase of the absorbed fraction of bulk ions.
David Terranova et al 2024 Nucl. Fusion
The RFX-mod2 installation is planned to be completed by 2024 and the start of operations is expected in 2025. The high flexibility of the machine (already tested in the previous RFX-mod experiment) allows operation in Reversed Field Pinch and tokamak configuration as well as ultra-low q pulses. In this work we present predictive analysis on transport, performances and plasma control in RFX-mod2 in view of the first experimental campaigns.
Bojiang Ding et al 2024 Nucl. Fusion
In order to sustain good LHW (lower hybrid wave) -plasma coupling for long pulse plasma operation, 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 such feedback control can work 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. Experiments between SMBI puffing and the gas-puffing fed by piezoelectric valve near the antenna are further investigated, showing that the response time on RC 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 such feedback control is effective for long pulse LHW-plasma coupling and the gas puffing by SMBI in the electric drift side of LHW antenna could offers an effective way to sustain good LHW coupling in steady state operation in future. Further optimization will be continued later.