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Gas puff imaging of plasma turbulence in the magnetic island scrape-off layer of W7-X
Nuclear Materials and Energy• Gas puff imaging of plasma turbulence in the magnetic island scrape-off layer on W7-X. • Observation of a radial profile of sheared poloidal flows and its relationships to the connection length profile. • Negligible level of radial transport in the confin…
Testing tungsten plasma facing components in WEST and AUG tokamaks
Nuclear FusionAbstract Next step fusion devices will face unprecedented heat loads and particle fluence with thousands of hours of plasma exposure on plasma-facing components (PFCs) cumulated over the entire lifetime of the device. These components must guarantee an acce…
A quasi-isodynamic stellarator configuration towards a fusion power plant
Journal of Plasma PhysicsThis work demonstrates that magnetohydrodynamic (MHD) stable, quasi-isodynamic (QI) stellarator equilibria with reduced turbulence can be generated with an optimised coilset. We present one such equilibrium which, when being generated by coils, maintains th…
Preliminary engineering design of the first wall and X-point divertor of the Helios planar coil stellarator
Fusion Engineering and Design“Helios” is a two-field-period, quasi-axisymmetric (“QA”) planar coil stellarator power plant that features a novel, toroidally-continuous, non-resonant X-point divertor isomorphic to that of a tokamak with resonant magnetic perturbations. This work describ…
Investigating island divertor physics with an extended stellarator two-point model
Nuclear FusionAbstract The two-point model by Stangeby has been widely used to gain insight into the relevant transport processes in the tokamak scrape-off-layer (SOL). It has been adapted into a stellarator two-point model (STPM) by Feng to consider additional perpendic…
Divertor shaping with neutral baffling as a solution to the tokamak power exhaust challenge
Communications PhysicsExhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping…
Magnetic mesh generation and field line reconstruction for scrape-off layer and divertor modeling in stellarators
Plasma Physics and Controlled FusionAbstract The design of divertor targets and baffles for optimal heat and particle exhaust from magnetically confined fusion plasmas requires a combination of fast, low-fidelity models (such as EMC3-Lite (Feng 2022 Plasma Phys. Control. Fusion 64 125012)) fo…
Overview of the third JET deuterium-tritium campaign
Plasma Physics and Controlled FusionAbstract JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET’s 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating…
SOLPS-ITER modelling of helium transport, recycling and pumping at the ASDEX Upgrade tokamak
Nuclear FusionAbstract Extrapolating the observed behavior of helium exhaust in current tokamaks towards future reactors requires the understanding of the underlying physical mechanisms determining helium transport, recycling and pumping. Helium compression is the main p…
Ex situ investigation of boron, hydrogen, and oxygen distributions on W7-X divertor surfaces using picosecond-LIBS under vacuum
KITopenQuantitative, spatially resolved measurements of near-surface material composition on plasma-facing components (PFCs) are critical for interpreting plasma–wall interaction (PWI) driven impurity sources, assessing wall-conditioning performance, and providing…
Ensuring the structural integrity of tokamak fusion power plants: challenges, progress and pathway
Philosophical Transactions of the Royal Society A Mathematical Physical and Engineering SciencesStructural integrity for fusion is an integrated multi-disciplinary subject spanning the science of materials, technology, engineering, health monitoring and simulation methods and algorithms for scrutinizing the assurance of reliable fusion reactor perform…
A fast helium beam diagnostic at the island divertor of W7-X for the investigation of plasma edge dynamics
Repository KITopen (Karlsruhe Institute of Technology)Understanding and quantifying particle and energy transport at the plasma edge region is crucial for magnetic confinement fusion. For this purpose, a new polychromator system with 1 µs time resolution was installed for the thermal helium beam diagnostic in…
On the minimum size of tokamaks aimed at fusion energy gain with optimum plasma temperature
Fusion Engineering and DesignThis work explores factors affecting the size reduction of tokamaks aiming at a specific thermal deuterium-tritium fusion energy gain Q f u s , using 1/2D physics models with IPB98(y,2) scaling law. A simplified Q f u s expression reveals an optimum plasma…
Regulation compliant AI for fusion: explainable image-based feedback control of divertor detachment in DIII-D tokamak
Nuclear FusionAbstract While artificial intelligence (AI) has been promising for fusion control, its inherent black-box nature will make compliant implementation in regulatory environments a challenge. This study implements and validates a real-time AI-enabled linear and…
Development of a code interface for coupled turbulence-transport simulations of tokamak edge plasmas
Plasma Physics and Controlled FusionAbstract The self-consistent simulation of the edge plasma is crucial for exploring the edge plasma solution compatible with high-performance plasma. While self-consistent edge plasma simulation is subject to the large gap between the turbulence and transpo…
Harnessing Toroidal Neutral Flows to Enhance Divertor Particle Exhaust
arXivIn 1991 Reiter et al. (1991 Plasma Phys. Control. Fusion 33 1579) considered the onerous exhaust requirements of ITER, and wrote: "The vacuum pumping problem of a fusion reactor will probably require some novel solution". Here we show that a toroidally orie…
Sensitivity Studies of Power Deposition Computed on Tokamak First Wall
IEEE Transactions on Plasma ScienceProfiles of power deposition on the first wall of the Joint European Torus (JET) tokamak experiment are fit by tracing an analytic representation for the distribution of power from midplane along lines of magnetic field. The technique is used to help design…
Demonstration of Super-X divertor exhaust control for transient heat load management in compact fusion reactors
Nature EnergyAbstract Nuclear fusion could offer clean, abundant energy. However, managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge. This is compounded in emerging compact reactor designs promising more cost-effe…
Core-edge integrated predictive studies of ST40 and NSTX plasmas with the scrape-off layer box model
Physics of PlasmasThe ability to model the interplay between the core and edge of tokamak plasmas is crucial to designing both the plasma operating scenario of a fusion pilot plant and the design of the tokamak itself. Scrape-off-layer (SOL) models that are tailored to integ…
Edge-localized mode mitigation enabled by active control of pedestal density gradient with new EAST tokamak divertor
Physics of PlasmasMitigation of large edge-localized modes (ELMs) has been achieved by actively reducing the pedestal density gradient with the EAST new right-angled lower divertor through changing the strike point position from the vertical target to the horizontal target.…
Modeling of the Sputtering Rate of the Tokamak First Wall by Atoms of Hydrogen Isotopes in the Near-Wall Plasma
Physics of Atomic NucleiAbstract A method for calculating sputtering of the tokamak first wall by fast charge-exchange neutral atoms of hydrogen isotopes in the near-wall plasma has been proposed. The application of the method has been shown by the example of sputtering of the ass…
Physics drivers for the plasma-facing component design of the COMPASS-U tokamak
Plasma Physics and Controlled FusionAbstract The main drivers for the metallic plasma-facing component (PFC) design of the COMPASS Upgrade (COMPASS-U) tokamak ( B T ⩽ 5 T, I P ⩽ 2 MA, R geo = 0.9 m, a geo = 0.27 m, t flattop ⩽ 3 s) derived from physics considerations are presented in this pap…
Thermo-structural assessment of the limiter inboard first wall design of the Divertor Tokamak Test facility
Fusion Engineering and DesignThe Divertor Tokamak Test facility (DTT) aims at investigating integrated power exhaust solutions that can be relevant for DEMO and future power plants. Such an ambitious goal imposes several constraints on the engineering design of the actively cooled plas…
Overview of the studies towards a Plasma Position Reflectometry system for Divertor Test Tokamak (DTT)
Journal of InstrumentationAbstract Plasma Position Reflectometry (PPR) is taking an important role in next-generation fusion machines, such as DEMO, as a diagnostic to monitor the position and shape of the plasma, complementing magnetic diagnostics. The Divertor Test Tokamak (DTT) f…