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.
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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Sehila M. Gonzalez de Vicente et al 2022 Nucl. Fusion 62 085001
Q.M. Hu et al 2024 Nucl. Fusion 64 046027
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.
I.A.M. Datta et al 2024 Nucl. Fusion 64 066016
The FuZE sheared-flow-stabilized Z pinch at Zap Energy is simulated using whole-device modeling employing an axisymmetric resistive magnetohydrodynamic formulation implemented within the discontinuous Galerkin WARPXM framework. Simulations show formation of Z pinches with densities of approximately 1022 m−3 and total DD fusion neutron rate of 107 per µs for approximately 2 µs. Simulation-derived synthetic diagnostics show peak currents and voltages within 10% and total yield within approximately 30% of experiment for similar plasma mass. The simulations provide insight into the plasma dynamics in the experiment and enable a predictive capability for exploring design changes on devices built at Zap Energy.
J. Elbez-Uzan et al 2024 Nucl. Fusion 64 037001
The discussion in the international community on how fusion power plants (FPPs) will be licenced and regulated is ongoing. As such, there is a concerted drive from the European stakeholders to understand the requirements from such a framework and how to best establish it with the aim of easing the licensing process of FPPs. Initiated by the EUROfusion consortium, a group of European experts were convened to produce a set of recommendations on the regulatory framework for the safety and licensing of FPPs. To do so effectively, the group assessed lessons learned from existing fusion facilities, reports by International Atomic Energy Agency and European Commission on FPP safety and the on-going work by the UK government, US Nuclear Regulatory Commission and Canadian Nuclear Safety Commission, as well as the licensing process of ITER. As a result, commonalities between fusion and fission were identified in terms of fundamental safety objectives which could facilitate parity in certain framework aspects. However, significant differences to any such implementation were also identified, particularly with respect to the lower hazard potential inherent to FPPs and how to remain proportionate to the associated safety challenges and the physical principles behind these two types of reactors together with their associated technologies. The recognition of the differences in the safety challenges in FPPs and fission-based nuclear power plants (NPPs) is paramount to future regulatory framework development. Ultimately, regulatory frameworks depend upon a country's legal framework, therefore it is apparent that a common global regulatory framework for FPPs is not possible. However, as with present-day NPP regulation, efforts could be made to develop harmonised approaches to FPP regulation to provide common levels of protection. In view of this objective, 12 recommendations are presented across 4 topics: regulations, international databases, codes and standards, safety demonstration rules and regulatory approaches. These recommendations are provided to inform and advise potential future actions on FPP regulatory framework and licencing process principles.
Semin Joung et al 2024 Nucl. Fusion 64 066038
A neural network, BES-ELMnet, predicting a quasi-periodic disruptive eruption of the plasma energy and particles known as edge localized mode (ELM) onset is developed with observed pedestal turbulence from the beam emission spectroscopy system in DIII-D. BES-ELMnet has convolutional and fully-connected layers, taking two-dimensional plasma fluctuations with a temporal window of size 128 µs and generating a scalar output which can be interpreted as a probability of the upcoming ELM onset. As approximately labeled inter-ELM broadband () fluctuations are given to the network, BES-ELMnet learns by itself ELM-related precursors arising before the onsets through supervised learning scheme. BES-ELMnet achieves the gradually increasing ELM onset probabilities between two consecutive ELMs during the inter-ELM phases and can forecast the first ELM onsets which occur after the high confinement mode transition. We further investigate the network generality in terms of the selected frequency band to ensure the use of BES-ELMnet for various operation regimes without changing the trained architecture. Therefore, our novel prediction method will enhance a proactive high confinement mode control of fusion-grade plasmas.
J.F. Parisi et al 2024 Nucl. Fusion 64 054002
A theoretical model is presented that for the first time matches experimental measurements of the pedestal width-height Diallo scaling in the low-aspect-ratio high-β tokamak NSTX. Combining linear gyrokinetics with self-consistent pedestal equilibrium variation, kinetic-ballooning, rather than ideal-ballooning plasma instability, is shown to limit achievable confinement in spherical tokamak pedestals. Simulations are used to find the novel Gyrokinetic Critical Pedestal constraint, which determines the steepest pressure profile a pedestal can sustain subject to gyrokinetic instability. Gyrokinetic width-height scaling expressions for NSTX pedestals with varying density and temperature profiles are obtained. These scalings for STs depart significantly from that of conventional aspect ratio tokamaks.
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.
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.
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.
Ting Long et al 2024 Nucl. Fusion 64 064002
We report on comprehensive experimental studies of turbulence spreading in edge plasmas. These studies demonstrate the relation of turbulence spreading and entrainment to intermittent convective density fluctuation events or bursts (i.e. blobs and holes). The non-diffusive character of turbulence spreading is thus elucidated. The turbulence spreading velocity (or mean jet velocity) manifests a linear correlation with the skewness of density fluctuations, and increases with the auto-correlation time of density fluctuations. Turbulence spreading by positive density fluctuations is outward, while spreading by negative density fluctuations is inward. The degree of symmetry breaking between outward propagating blobs and inward propagating holes increases with the amplitude of density fluctuations. Thus, blob-hole asymmetry emerges as crucial to turbulence spreading. These results highlight the important role of intermittent convective events in conveying the spreading of turbulence, and constitute a fundamental challenge to existing diffusive models of spreading.
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M. Sato et al 2024 Nucl. Fusion 64 076021
Effects of the kinetic thermal ions (KTIs) on ideal infernal modes and resistive infernal modes have been investigated by using magnetohydrodynamic (MHD) simulation without KTIs and kinetic-MHD hybrid simulation with KTIs. For the ideal infernal modes, the pressure profile is significantly flattened at the saturated state for both the models with and without the KTIs. As the beta value decreases, the ideal infernal modes are stabilized while the resistive infernal modes are still unstable. For the resistive infernal modes, while the saturated pressure profile is significantly flattened in the MHD simulation without KTIs, the pressure profile is not flattened at the saturated state in the kinetic-MHD hybrid simulation with KTIs. The suppression of the saturation level by the effects of the KTIs results from the phase mismatch between the radial velocity and perturbed pressure mode structures. This indicates that KTIs play an essential role for the suppression of pressure profile flattening due to slowly growing resistive MHD instabilities.
Lan Yin et al 2024 Nucl. Fusion 64 076020
Efficient ion heating is crucial for future fusion devices, and the only way to heat ions directly is ion cyclotron resonance heating. 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 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.
M. Morbey et al 2024 Nucl. Fusion 64 076019
The vapor-box, a liquid metal design for the divertor, utilizes lithium recirculation through evaporation and condensation. Safety concerns arise from Li-D/T formation and co-deposition on vapor-box walls and the first wall, affecting tritium retention. Additively manufactured tungsten capillary porous structure (CPS) samples with Li were exposed to high heat flux D plasmas in the linear plasma device Magnum-PSI, to study D retention in Li–D co-deposition dependence on substrate temperature (200 C–428 C) and distance from the plasma beam center (25–85 mm). The D:Li ratio was determined via in-situ ion beam diagnostics with simultaneously analyzed Nuclear Reaction Analysis and Elastic Backscattering Spectroscopy spectra to maximize the precision. Experimental results approach close to the theoretical maximum at 40:60 D:Li ratio and deposited film thickness ranging from 0.02 to 3.2 µm. Witness plate temperatures above 400 C yielded Li films under 150 nm in thickness with lower D:Li ratios (5:95 D:Li ratio). At this temperature LiD decomposition pressure is comparable with vessel pressure during plasma. SOLPS-ITER simulations narrowed CPS surface temperature to 650 C–700 C, indicating Li+ plasma dominance near the target surface. Redeposition ratio of lithium on the CPS surface was determined to be around 80, matching quartz crystal microbalance results. However, SOLPS-ITER simulations lacked accuracy in recreating observed Li and D deposition layers on WPs, improvements are needed to model plasmas with significant Li quantities. Extension of SOLPS-ITER simulations to include LiD molecules and enhance heat flux accuracy is crucial for better alignment with experimental data.
M. Rud et al 2024 Nucl. Fusion 64 076018
Tomographic reconstructions of a 3D fast-ion constants-of-motion phase-space distribution function are computed by inverting synthetic signals based on projected velocities of the fast ions along the diagnostic lines of sight. A spectrum of projected velocities is a key element of the spectrum formation in fast-ion D-alpha spectroscopy, collective Thomson scattering, and gamma-ray and neutron emission spectroscopy, and it can hence serve as a proxy for any of these. The fast-ion distribution functions are parameterised by three constants of motion, the kinetic energy, the magnetic moment and the toroidal canonical angular momentum. The reconstructions are computed using both zeroth-order and first-order Tikhonov regularisation expressed in terms of Bayesian inference to allow uncertainty quantification. In addition to this, a discontinuity appears to be present in the solution across the trapped-passing boundary surface in the three-dimensional phase space due to a singularity in the Jacobian of the transformation from position and velocity space to phase space. A method to allow for this apparent discontinuity while simultaneously penalising large gradients in the solution is demonstrated. Finally, we use our new methods to optimise the diagnostic performance of a set of six fans of sightlines by finding where the detectors contribute most complementary diagnostic information for the future COMPASS-Upgrade tokamak.
T. Stoltzfus-Dueck and R. Brzozowski III 2024 Nucl. Fusion 64 076017
Using the assumption of a weak normalized turbulent viscosity, usually valid in practice, the modulated-transport model (Stoltzfus-Dueck 2012 Phys. Plasmas19 055908) is generalized to allow the turbulent transport coefficient to vary in an arbitrary way on radial and poloidal position. The new approach clarifies the physical interpretation of the earlier results and significantly simplifies the calculation, via a boundary-layer asymptotic method. Rigorous detailed appendices verify the result of the simple boundary-layer calculation, also demonstrating that it achieves the claimed order of accuracy and providing a concrete prediction for the strong plasma flows in the immediate vicinity of the last closed flux surface. The new formulas are used to predict plasma rotation at the core-edge boundary, in cases with and without externally applied torque. Dimensional formulas and extensive discussion are provided, to support experimental application of the new model.
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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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Imadera et al
Aiming at a fuel supply through particle pinch effects, turbulent particle transport is studied by gyrokinetic flux-driven ITG/TEM simulations. It is found that ITG/TEM turbulence can drive ion particle pinch by E×B drift (n≠0) when the ion temperature gradient is steep enough. Electron particle pinch is also driven by E×B drift (n≠0) in the case with the steep electron temperature gradient. Such an electron particle pinch can trigger an ambipolar electric field, leading to additional ion particle pinch by not only magnetic drift but also E×B drift (n=0). These results suggest that a density peaking of bulk ions due to turbulent fluctuations can be achieved by sufficiently strong both ion and electron heating.
Zlobinski et al
The paper reports the first demonstration of in situ LID-QMS application on a large scale fusion device performed in summer 2023. LID-QMS allows direct measurements of the fuel inventory of plasma facing components without retrieving them from the fusion device. The diagnostic desorbs the retained gases by heating a 3 mm diameter spot on the wall using a 1 ms long laser pulse and detects them by Quadrupole Mass Spectrometry (QMS). Thus, it can measure the gas content at any wall position accessible to the laser. The successful LID-QMS application in laboratory scale and on medium size fusion devices has now been demonstrated on the larger scale and it is already foreseen as tritium monitor diagnostic in ITER. This in situ diagnostic gives direct access to retention physics on a short timescale instead of campaign-integrated measurements and can assess the space-resolved efficacy of detritation methods. LID-QMS can be applied on many materials: on Be deposits like in JET, B deposits like in TEXTOR, C based materials or on bulk-W.
Li et al
The safe operation of most tokamaks, especially the large ones, rely on the feedback control of the vertical displacement events (VDEs). However, most these feedback control systems are based on the axisymmetric VDE models. In this work, we use NIMROD simulations to study the role of non-axisymmetric perturbations in free drift vertical displacement events on EAST. The high-$n$ modes in non-axisymmetric VDE grow first, which drive the formation of high-$n$ magnetic island chains. Subsequently, the magnetic island chains grow and overlap with each other, leading to the destruction of the magnetic flux surface, which induces a minor disruption and accelerates the start of the following major disruption. The magnetic island and the stochastic magnetic field allow the toroidally non-axisymmetric poloidal plasma current to jet towards the hoop force direction, forming the finger-like and filamentary structures. Such a plasma current non-axisymmetry strongly depends on the anisotropy in thermal transport coefficients.
Schmid et al
Future fusion reactors will have to breed enough tritium (T) to sustain continuous operation and to produce excess T to power up other fusion reactors. Therefore, T is a scarce resource that must not be lost inside the fusion power plants systems. The factor that describes the T production is the "tritium breeding ratio" (TBR) which is the ratio of the breading rate in atoms per second to the burn rate in atoms per second. Its value is calculated from neutronics analyses of the breeding process in the blanket and coupled dynamics of the T processing plant. However, these calculations generally ignore the T transport and loss in the first wall by assuming essentially instantaneous recycling of the impinging T in-flux. In this paper the transport and retention of T in the main chamber first wall of a future EU-DEMO reactor is investigated based on the available material data and expected particle loads onto the wall. Two breeding blanket concepts are compared WCLL (Water Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) and the resulting wall-loss probabilities are compared with a simple balance model that describes the maximum allowable wall loss given a TBR to achieve T-self-sufficiency.
Wen et al
The role of a series of quasi-coherent modes (QCMs) on the turbulence properties has been investigated during Type-III ELMs under HL-2A high confinement mode (H-mode) scenarios. The QCMs are essentially electrostatic and appear during the inter-ELM periods, with a frequency ranging from 20 kHz to 60 kHz. These QCMs are localized in the pedestal region and are related to the saturation of density gradient in the pedestal. Nonlinear couplings between QCMs and ambient turbulence have been observed and also verified through the envelope modulation of turbulence in density by the radial electrical field fluctuation of the modes. The presence of QCMs can increase the radial and poloidal turbulence correlation lengths, thereby modulating the turbulent transport. Experimental results show that QCMs significantly impact pedestal turbulence and transport by increasing the correlation length as well as the decorrelation time of turbulent eddies. The flow shearing rate in pedestal region is also enhanced to a level that surpasses the decorrelation frequency of turbulence, thus, the existence of QCMs has the ability to put off the ELM burst due to relative stronger stabilization of turbulence by sheared flows. The above results have demonstrated that the pedestal dynamics is largely determined by the complex interactions among QCMs, turbulence and the shear flow.
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Kenji Imadera et al 2024 Nucl. Fusion
Aiming at a fuel supply through particle pinch effects, turbulent particle transport is studied by gyrokinetic flux-driven ITG/TEM simulations. It is found that ITG/TEM turbulence can drive ion particle pinch by E×B drift (n≠0) when the ion temperature gradient is steep enough. Electron particle pinch is also driven by E×B drift (n≠0) in the case with the steep electron temperature gradient. Such an electron particle pinch can trigger an ambipolar electric field, leading to additional ion particle pinch by not only magnetic drift but also E×B drift (n=0). These results suggest that a density peaking of bulk ions due to turbulent fluctuations can be achieved by sufficiently strong both ion and electron heating.
Miroslaw Zlobinski et al 2024 Nucl. Fusion
The paper reports the first demonstration of in situ LID-QMS application on a large scale fusion device performed in summer 2023. LID-QMS allows direct measurements of the fuel inventory of plasma facing components without retrieving them from the fusion device. The diagnostic desorbs the retained gases by heating a 3 mm diameter spot on the wall using a 1 ms long laser pulse and detects them by Quadrupole Mass Spectrometry (QMS). Thus, it can measure the gas content at any wall position accessible to the laser. The successful LID-QMS application in laboratory scale and on medium size fusion devices has now been demonstrated on the larger scale and it is already foreseen as tritium monitor diagnostic in ITER. This in situ diagnostic gives direct access to retention physics on a short timescale instead of campaign-integrated measurements and can assess the space-resolved efficacy of detritation methods. LID-QMS can be applied on many materials: on Be deposits like in JET, B deposits like in TEXTOR, C based materials or on bulk-W.
Haolong Li et al 2024 Nucl. Fusion
The safe operation of most tokamaks, especially the large ones, rely on the feedback control of the vertical displacement events (VDEs). However, most these feedback control systems are based on the axisymmetric VDE models. In this work, we use NIMROD simulations to study the role of non-axisymmetric perturbations in free drift vertical displacement events on EAST. The high-$n$ modes in non-axisymmetric VDE grow first, which drive the formation of high-$n$ magnetic island chains. Subsequently, the magnetic island chains grow and overlap with each other, leading to the destruction of the magnetic flux surface, which induces a minor disruption and accelerates the start of the following major disruption. The magnetic island and the stochastic magnetic field allow the toroidally non-axisymmetric poloidal plasma current to jet towards the hoop force direction, forming the finger-like and filamentary structures. Such a plasma current non-axisymmetry strongly depends on the anisotropy in thermal transport coefficients.
Klaus Schmid et al 2024 Nucl. Fusion
Future fusion reactors will have to breed enough tritium (T) to sustain continuous operation and to produce excess T to power up other fusion reactors. Therefore, T is a scarce resource that must not be lost inside the fusion power plants systems. The factor that describes the T production is the "tritium breeding ratio" (TBR) which is the ratio of the breading rate in atoms per second to the burn rate in atoms per second. Its value is calculated from neutronics analyses of the breeding process in the blanket and coupled dynamics of the T processing plant. However, these calculations generally ignore the T transport and loss in the first wall by assuming essentially instantaneous recycling of the impinging T in-flux. In this paper the transport and retention of T in the main chamber first wall of a future EU-DEMO reactor is investigated based on the available material data and expected particle loads onto the wall. Two breeding blanket concepts are compared WCLL (Water Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) and the resulting wall-loss probabilities are compared with a simple balance model that describes the maximum allowable wall loss given a TBR to achieve T-self-sufficiency.
Jie Wen et al 2024 Nucl. Fusion
The role of a series of quasi-coherent modes (QCMs) on the turbulence properties has been investigated during Type-III ELMs under HL-2A high confinement mode (H-mode) scenarios. The QCMs are essentially electrostatic and appear during the inter-ELM periods, with a frequency ranging from 20 kHz to 60 kHz. These QCMs are localized in the pedestal region and are related to the saturation of density gradient in the pedestal. Nonlinear couplings between QCMs and ambient turbulence have been observed and also verified through the envelope modulation of turbulence in density by the radial electrical field fluctuation of the modes. The presence of QCMs can increase the radial and poloidal turbulence correlation lengths, thereby modulating the turbulent transport. Experimental results show that QCMs significantly impact pedestal turbulence and transport by increasing the correlation length as well as the decorrelation time of turbulent eddies. The flow shearing rate in pedestal region is also enhanced to a level that surpasses the decorrelation frequency of turbulence, thus, the existence of QCMs has the ability to put off the ELM burst due to relative stronger stabilization of turbulence by sheared flows. The above results have demonstrated that the pedestal dynamics is largely determined by the complex interactions among QCMs, turbulence and the shear flow.
Naoto Tsujii et al 2024 Nucl. Fusion
Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8\,kA was achieved. Coupled ray-tracing and Fokker-Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.
Paola Mantica et al 2024 Nucl. Fusion
In the JET DTE2 campaign a new method was successfully tested to detect the heating of bulk electrons by α-particles, using the dynamic response of the electron temperature Te to the modulation of Ion Cyclotron Resonance Heating (ICRH). A fundamental Deuterium (D) ICRH scheme was applied to a Tritium-rich Hybrid plasma with D-NBI. The modulation of the ion temperature Ti and of the ICRH accelerated Deuterons leads to modulated α-heating with a large delay with respect to other modulated electron heating terms. A significant phase delay of ~40º is measured between central Te and Ti, which can only be explained by α-particle heating. Integrated modelling using different models for ICRH absorption and ICRH/NBI interaction reproduces the effect qualitatively. Best quantitative agreement with experiment is obtained with the Europen Transport Solver/Heating and Current Drive workflow.
M. Sato et al 2024 Nucl. Fusion 64 076021
Effects of the kinetic thermal ions (KTIs) on ideal infernal modes and resistive infernal modes have been investigated by using magnetohydrodynamic (MHD) simulation without KTIs and kinetic-MHD hybrid simulation with KTIs. For the ideal infernal modes, the pressure profile is significantly flattened at the saturated state for both the models with and without the KTIs. As the beta value decreases, the ideal infernal modes are stabilized while the resistive infernal modes are still unstable. For the resistive infernal modes, while the saturated pressure profile is significantly flattened in the MHD simulation without KTIs, the pressure profile is not flattened at the saturated state in the kinetic-MHD hybrid simulation with KTIs. The suppression of the saturation level by the effects of the KTIs results from the phase mismatch between the radial velocity and perturbed pressure mode structures. This indicates that KTIs play an essential role for the suppression of pressure profile flattening due to slowly growing resistive MHD instabilities.
Chengkang Pan 2024 Nucl. Fusion
A new mechanism for the high-Z impurity neoclassical particle transport in the tokamak plasmas due to the single-null divertor configuration is discovered for the first time. It will play an important role in the tokamak plasma pedestal region with the coexisting of the strong bulk ion radial gradients and the up/down asymmetry of the poloidal magnetic field. The outward (inward) high-Z impurity neoclassical particle transport will be driven with the B×▽B drift towards (away from) the X-point. The new finding indicates that the International Thermonuclear Experimental Reactor (ITER) lower single-null divertor configuration will be beneficial for screening and flushing out the high-Z impurity in the pedestal region with the toroidal magnetic field in the normal direction ( B×▽B drift towards the X-point).
Koki Imada et al 2024 Nucl. Fusion
The first pedestal stability and structure analysis on the new MAST Upgrade (MAST-U) spherical tokamak H-mode plasmas is presented. Our results indicate that MAST-U pedestals are close to the low toroidal mode number (n) peeling branch of the peeling-ballooning instability, in contrast with MAST H-mode pedestals which were deeply in the high-n ballooning branch. This offers the possibility of reaching the ELM-free quiescent H-mode [1] or high-performance super H-mode [2,3] regimes. In addition, the coupling between the peeling and ballooning branches is weak in MAST-U, suggesting that a path to very high pedestal pressure gradient at high density may exist with sufficient heating power. A possible explanation for the differences between MAST and MAST-U pedestal stability is given in terms of plasma shaping parameters, in particular squareness and elongation, as well as the pedestal top temperature and collisionality.
[1] K.H. Burrell et al, PPCF 2005
[2] P.B. Snyder et al, NF 2015
[3] P.B. Snyder et al, NF 2019