ITER Plasma Research Articles

The consequences of tritium mix for simulated ion cyclotron emission spectra from deuterium-tritium plasmas

Abstract Measurements of ion cyclotron emission (ICE) are obtained from most large magnetically confined fusion (MCF) plasma experiments, and may be used in future to quantify properties of the fusion-born alpha-particle population in deuterium-tritium (DT) plasmas in ITER. ICE is driven by spatially localised, strongly non-Maxwellian, minority energetic ion populations which relax collectively under the magnetoacoustic cyclotron instability (MCI). ICE spectral peaks are typically observed at, near, or separated by, integer harmonics of the energetic ion cyclotron frequency. Here, for the first time, we study how simulated ICE spectra from DT plasmas vary with tritium concentration and compare with an observed JET DT ICE spectrum. We incorporate a population of thermal tritons, in addition to thermal deuterons and an energetic minority population of fusion-born alpha-particles, in simulations with the kinetic particle-in-cell (PIC) code EPOCH. This code has previously been used extensively for interpretation of ICE observations in terms of the self-consistent gyro-resolved collective Maxwell-Lorentz dynamics of tens of millions of simulation particles. Our simulation parameters are relevant to the ICE-generative outer midplane edge region of JET DT plasma 26148, which included 11% tritium. Quantifying the variation of simulated ICE power spectra with tritium concentration reveals that our simulation with 11% tritium concentration most accurately represents the observed ICE spectrum from this plasma. This outcome is encouraging for the diagnostic application of ICE to fusion plasmas containing two thermal ion species, including future DT plasmas.

Simulations of the stationary Q = 10 and the exit phase from the flat-top of an ITER 15MA baseline scenario: predictive JINTRAC simulation with a consistent treatment of D and T in the whole plasma

Abstract Designing a robust termination scenario for a burning ITER plasma is a challenge that requires extensive core plasma and divertor modelling. The presented work consists of coupled core/edge/SOL/divertor simulations, performed with the JINTRAC code, to study the Q = 10 flat-top phase and exit phase of the ITER 15 MA/5.3 T DT scenario. The modelling utilizes the recently implemented option to treat deuterium and tritium separately in the SOL/divertor, enabling a consistent treatment of deuterium and tritium in the whole plasma volume, which is a unique capability of JINTRAC. In addition, these are the first JINTRAC simulations of this scenario that use a first-principles transport model to self-consistently model the ECRH power deposition and to include tungsten while keeping track of tungsten sputtering and accumulation. The flat-top simulations demonstrate the possibility of sustaining a steady state fusion Q of 10 using pure deuterium gas puffs together with DT mixed pellets, which is an option to make a more effective use of tritium. Simulations of the exit phase are set up sequentially, with each phase providing initial conditions for the next, starting with a density decay at full current and auxiliary power, and demonstrate the possibility of reducing the density robustly within a few seconds. Following the density decay, a subsequent auxiliary power ramp-down in H-mode is performed with a late H–L transition at low auxiliary power, which may provide an option for the optimization of the plasma termination. The final ramp-down phase consists of a current ramp-down in L-mode to 3.75 MA.

Development of ITER first wall heat load feedback control

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

Impact of repetitive ELM transients on ITER divertor tungsten monoblock top surfaces

Abstract Owing to the high stored energy of ITER plasmas, the heat pulses due to uncontrolled Type I edge localized modes (ELMs) can be sufficient to melt the top surface of several poloidal rows of tungsten monoblocks in the divertor strike point regions. Coupled with the melt motion associated with tungsten in the strong tokamak magnetic fields, the resulting surface damage after even a comparatively small number of such repetitive transients may have a significant impact on long-term stationary power handling capability. The permissible numbers set important boundaries on operation and on the performance required from the plasma control system. Modelling is carried out with the recently updated MEMENTO melt dynamics code, which is tailored to tackle melt motion problems characterized by a vast spatio-temporal scale separation. The crucial role of coupling between surface deformation and shallow angle heat loading in aggravating melt damage is highlighted. As a consequence, the allowable operational space in terms of ELM-induced transient heat loads is history-dependent and once deformation has occurred, weaker heat loads, incapable of melting a pristine surface, can further extend the damage.

Overview of the EUROfusion Tokamak Exploitation programme in support of ITER and DEMO

Abstract Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called "Tokamak Exploitation”. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

ATEP: an advanced transport model for energetic particles

In this paper we report on the implementation and verification of a phase-space resolved energetic particle (EP) transport model. It is based on a first-principle theoretical framework, i.e. the system of non-linear gyrokinetic equations and the related transport equations. Its focus is primarily directed toward understanding the meso-scale character of EPs and its consequences. Compared to the conventional description of thermal radial transport via a one-dimensional radial diffusion equation, the newly developed model is three-dimensional using canonical constants-of-motion (CoM) variables. The model does not assume diffusive processes to be dominant a priori, instead the EP fluxes are self-consistently calculated and directly evolved in CoM space. We use the EP-Stability workflow and the HAGIS code to determine the phase space fluxes explicitly either in the limit of constant mode amplitudes or an energy-conserving quasi-linear model. As an application of the model the transport of neutral-beam-generated EPs due to a toroidal Alfvén eigenmode in an ITER plasma is investigated. As there are no sources and collisions taken into account so far (for an extension of the model see the companion paper (Meng et al 2024 Nucl. Fusion accepted)), the results cannot be considered as an exhaustive study, but rather as a practical demonstration of the conceptual framework on the way to a comprehensive reduced description of burning plasmas.

Inter-ELM pedestal turbulence dynamics dependence on q95 and temperature gradient

A series of dedicated experiments from the DIII-D tokamak provide spatially and temporally resolved measurements of electron density and temperature, and multiscale and multichannel fluctuations over a wide range of conditions. Measurements of long wavelength density fluctuations in the type-I ELMing H-mode pedestals routinely reveal a coexistence of multiple instabilities that exhibit dramatic different dynamic behaviors as q95 and temperature gradients are varied, apparently responsible for limiting pedestal temperature profiles. Two distinct frequency bands of density fluctuations are modulated by an ELM cycle with frequency above 200 kHz propagating in the electron diamagnetic direction in the lab frame (electron mode) and below 200 kHz propagating in the ion diamagnetic direction (ion mode). The electron mode amplitude peaks near the electron temperature gradient region and increases with q95 which seems to be correlated with the increased χe at higher q95, similar to the characteristics expected for the micro-tearing mode (MTM). At higher q95, during the inter-ELM period, the ion mode decays at the later phase of the ELM cycle. Consistently, the poloidal correlation length of the ion mode is also found to reduce, which suggests the possible E × B flow shear suppression of the ion mode at the later phase of the ELM cycle as the Er well recovers. In contrast, the electron mode grows during the ELM cycle and reaches saturation at around 50%–60% of the ELM period. Linear gyrokinetic simulations find the MTMs to be the most unstable mode in the pedestal electron temperature gradient region. The higher q95 and lower magnetic shear destabilize the MTMs. These observations provide key insights into the underlying physics of multifield properties and a rich dataset of experimental ‘fingerprints’ that enable new tests of theoretical pedestal models and lead to the development of a predictive model for pedestal formation on the ITER and future burning plasma experiments.

Core electron temperature turbulence and transport during sawtooth oscillations in the DIII-D tokamak

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 T e inversion radius and/or safety factor q = 1 magnetic surface. The T e turbulence amplitude in the core during the sawtooth ramp exhibits a critical T e gradient behavior inside but not near the T e 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.

Strategy to systematically design and deploy the ITER plasma control system: A system engineering and model-based design approach

The paper details the process of developing the ITER Plasma Control System (PCS), that is, how to design and deploy it systematically, in the most efficient and effective manner. The integrated nature of the ITER PCS, with its multitude of coupled control functions, and its long-term development, calls for a different approach than the design and short-term deployment of individual controllers. It requires, in the first place, a flexible implementation strategy and system architecture that allows system re-configuration and optimization throughout its development. Secondly, a model-based system engineering approach is carried out, for the complete PCS development, i.e. both its design and deployment. It requires clear definitions for both the PCS role and its functionality, as well as definitions of the design and deployment process itself. The design and deployment process is shown to allow tracing the relationships of the many individual design and deployment aspects, such as system requirements, assumed operation use-cases and response models, and eventually verification and functional validation of the system design. The functional validation will make use of a dedicated PCS simulation platform that includes the description of the control function design as well as plant, actuator and sensor models that enable the simulation of these functions. By establishing a clear understanding of the interconnected steps involved in designing, implementing, commissioning, and operating the system, a more systematic approach is achieved. This ensures the completion of a comprehensive design that can be deployed efficiently, hence preventing the loss of precious operational time needed to debug and retune control functions and more importantly avoiding tokamak discharge disruptions.

Feasibility of a collective Thomson scattering diagnostic for burning plasma control on DEMO

Diagnostic systems are essential for burning plasma control and operation of DEMO, but are at the same time challenging in terms of design and integration. The harsh environment around the DEMO plasma, and the space restrictions and need to maximize the first-wall area used for tritium breeding, set limitations on the number and type of diagnostics to be installed. This will focus the efforts on diagnostics needed for control of the DEMO plasma. The robustness and versatility of a microwave-based Collective Thomson Scattering (CTS) diagnostic make it worthwhile to investigate the potential of a DEMO CTS diagnostic. We present the initial exploratory effort on this.The study builds on the experience of existing CTS experiments on e.g. ASDEX Upgrade, and on the recent development of the ITER CTS system, which focuses on measurements of fast ion dynamics in the burning ITER plasma. The initial target of the DEMO CTS diagnostic was to use an Electron Cyclotron Resonance Heating (ECRH) gyrotron beam as the probing source, with the receiving quasi-optical system being a dedicated CTS setup. Based on raytracing calculations including signal-to-noise estimates, it is found that such a setup is not viable. Here, we present studies of alternative solutions, including assessments of which DEMO plasma parameters the CTS diagnostic may contribute to determine.

Predictive modeling of Alfvén eigenmode stability in inductive scenarios in JT-60SA

The JT-60SA device offers unique conditions before ITER for the study of the interaction of energetic particles with plasma waves. With similar dimensions to JET, e.g., a major radius but with a slightly more elongated plasma volume, JT-60SA is used as a high-power device where additional heating power (including 10 MW of the 500 keV Neutral Beam Injection) of up to 41 MW and the potential for high non-inductive plasma current operation pave the path for numerous challenges in physics on MHD stability, in particular, when considering the effects of energetic particles. Several operational scenarios with ITER and DEMO-relevant plasma regimes, in terms of non-dimensional plasma parameters, are anticipated. In this work, the stability of Alfvén eigenmodes (AEs) in variants of two of the most relevant operational scenarios with single null is analyzed: a full Ip inductive scenario at high density (1.1 × 1020 m−3 on-axis electron density) and 5.48MA/2.05T toroidal plasma current and magnetic field, and an advanced (hybrid) scenario with an ion energy transport barrier (ITB) and 3.5MA/2.28T toroidal plasma current and magnetic field. The workflow included the CRONOS code to establish the scenario, the ASCOT code to calculate the slowing-down energetic particle distributions for a positive/negative ion source-based neutral beam, and the MISHKA/CASTOR-K suite to calculate the MHD spectra of AEs and the associated drive/damping contributions from the NBI energetic ions, as well as the thermal ion landau damping. The systematic analysis, over a large Fourier space of the toroidal mode number/mode frequency, provides evidence that although a significant fraction of supra-Alfvénic particles stemming from the negative ion source-based neutral beam (500 keV) can, in some cases, drive to AEs in both scenarios, it is not enough to overcome the thermal ion landau damping. In addition, the advanced scenario with ITB is shown to be stable against AEs localized in the vicinity of the barrier as well, offering good prospects of sustainability of the plasma performance and of ITB. Finally, some sensitivity scan results are shown on the influence of fast ion density and q-profile on the AE mode spectra and stability.

Error-field penetration thresholds in ohmically heated ITER and SPARC plasmas

The critical n = 1, n = 2, and n = 3 error-field amplitudes needed to trigger error-field penetration in ITER and SPARC, steady-state, ohmically heated plasmas are calculated using a standard asymptotic matching approach. The calculation incorporates plasma impurities, trapped particles, the bootstrap current, and neoclassical poloidal flow-damping. The energy confinement time is specified by the neo-Alcator scaling law in the low-density linear Ohmic confinement (LOC) regime and by the ITER-89P L-mode scaling law in the high-density saturated Ohmic confinement (SOC) regime. The response of the plasma in the inner region is calculated using a linearized version of the four-field model. At the normal operating electron number density, diamagnetic levels of rotation are found to be sufficient to protect ITER and SPARC ohmically heated plasmas from m=2/n=1 error-field penetration. On the other hand, SPARC, and especially ITER, ohmically heated plasmas may be vulnerable to n > 1 error-field penetration. ITER and SPARC ohmically heated plasmas are also slightly more susceptible to error-field penetration when the electron fluid at the rational surface rotates in the ion diamagnetic direction, rather than the electron diamagnetic direction. At electron number densities that are sufficiently low that the plasma lies in the LOC confinement regime, the error-field penetration threshold increases with increasing density. However, as soon as the electron number density becomes large enough that the plasma enters the SOC regime, the increase in the error-field penetration threshold with increasing density levels off.

Kinetic modelling of start-up runaway electrons in KSTAR and ITER

Understanding the formation of start-up runaway electrons (REs) is essential to ensure successful plasma initiation in ITER. The design of ITER start-up scenarios requires not only predictive simulations but also a validation of assumptions. The objective of this study is to strengthen the physical background required for predictive simulations aimed at ITER plasma start-up design, by validating the model assumptions. Through kinetic simulations, this study examines the validity of steady-state models for Dreicer generation under slowly-varying time scales relevant to plasma start-up and investigates the finite energy effect, commonly neglected, on the runaway avalanche growth rate. The research findings provide insights into situations where kinetic simulations are necessary. To secure a margin-of-control scheme without kinetic simulation, we suggest a strategy of scanning the Coulomb logarithm in fluid simulations as an alternative to predict runaway current takeover and avoid RE dominant scenarios. Ultimately, this paper seeks to offer a robust physical background, practically supporting the successful design of ITER start-up scenarios.

AC Losses Calculations for the ITER CS and PF Magnet Systems During Plasma Operation

While the ITER magnet system assembly is progressing, models are being developed to predict the performances of the coils in support of the preparation of commissioning and operation. In particular, the AC losses generated in the coils during a plasma scenario need to be estimated as input to thermo-hydraulic performance simulations. The largest AC losses during plasma operation are expected in the CS and PF coils. This paper discusses the results of simulations of AC losses in these coils during a representative ITER plasma pulse. These calculations also take into account the electromagnetic screening provided by the ITER passive structures, most importantly the vacuum vessel, discussing its impact on the conductor AC losses. The results will also be compared to earlier predictions. Finally, the CS and PF coil AC losses expected during commissioning will be briefly discussed.

Instrumented fast valve for the ITER DMS support laboratory test bench

The ITER Disruption Mitigation System (DMS) Support Laboratory at the Centre for Energy Research, Budapest, Hungary aims at the characterization of the de-sublimation, acceleration and shattering of the cryogenic pellets, which will be fired upon request into the ITER plasma to mitigate the detrimental consequences of disruptions. The test bench includes a fast-acting valve for pellet acceleration, which releases high-pressure gas on a millisecond time scale. The valve is a new development since no valve on the market existed which would have met the requirements. The electromechanical modelling of the valve actuator has been validated against a laboratory experiment, the final system was built and thoroughly tested on a test bench to characterize the valve properties, and to demonstrate that the system fulfils the requirements.

Evaluation of ITER divertor shunts as a synthetic diagnostic for detachment control

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control.

Integration of new sets of magnets for improved plasma confinement in the SPIDER experiment

SPIDER is the full-scale prototype of the plasma source of the negative-ion driven neutral beam injector for the heating and current drive of the ITER plasma. The uniqueness and complexity of the system requested this ad hoc test stand aiming at optimizing the performance of the RF inductively generated plasma, negative ion production and extraction, electron filtering, and robustness and controllability of all systems required to work together. After about three years of operation, presently SPIDER is in a long shutdown, in which the whole plasma source and accelerator were dismounted. In this phase, additional modifications with respect to the original design will be introduced to improve the system performance, driven by the experience acquired in the last years.These include the addition of further sets of permanent magnets in the plasma source expansion chamber and around the RF drivers, with the aim of improving the plasma confinement and consequently its density and possibly its uniformity. The present paper reports the study and the analyses behind this modification, which impacts on the original already complex magnetic configuration, made particularly difficult by the limited space available and the high voltages. The use of ferromagnetic shields, necessary to limit stray fields possibly increasing the breakdown probability, make the design particularly complex because of the greater impact on the previous configuration. An iterative process between analyses to determine the ideal configuration and CAD verifications was required. The analyses had to take into account the new magnetic configuration to be created in the particular area of interest, and the overall configuration in order to not compromise its efficacy.

Theoretical investigation of braking of tearing mode rotation by resistive walls in ITER

The locking of the 2/1 tearing mode to the resistive wall in the ITER tokamak (15 MA inductive scenario 2) is investigated theoretically using a cylindrical asymptotic matching model. The model takes into account the fact that ITER plasmas will effectively be surrounded by two walls; the inner blanket module layer with a time constant of about 23 ms, and the outer vacuum vessel with a time constant of about 380 ms. The model also takes cognizance of the fact that neither the blanket module layer nor the vacuum vessel can be accurately described as “thin” walls (in the ordinarily accepted sense). The model incorporates changes in both the plasma poloidal and the toroidal angular velocity profiles, in response to the electromagnetic braking torque that develops at the rational surface, because it turns out that neoclassical poloidal flow-damping is not strong enough to completely suppress changes in the poloidal velocity. Finally, the model accurately calculates changes in the poloidal and toroidal plasma angular velocity profiles by evolving the full angular equations of motion, taking the electromagnetic braking torque, plasma inertia, plasma viscosity, and poloidal flow-damping into account. The time required for the 2/1 tearing mode to grow from a small amplitude to a sufficient one to lock to the walls is found to be about 3.5 s. The critical full radial island width at which wall locking is triggered is found to be about 9% of the plasma minor radius.

A case study of the real-time framework for the implementation of the ITER plasma control system

ITER CODAC is in the process of developing a tailored software solution to carry out necessary control functions for tokamak operation. CODAC has successfully completed the final review of the implementation of the ITER Plasma Control System (PCS) for first plasma (FP), and formally commenced its production in accordance with the conversion mechanism of the Real-Time Framework (RTF) software suite. Rich functionality of RTF built-ins and its flexibility of reconfigurable applications facilitates the development and deployment of complex real-time applications such as commonly found within PCS and the ITER diagnostics.In the practical application of the framework, CODAC has prototyped PCS architecture components to verify how actual operation complies with the intended design. Prototyping initiates CODAC preparation for system validation and verification, and its commissioning. Agile production is based on the model presented in the PCS Simulation Platform (PCSSP). Several fundamental functions such as exception handling, quality tag controlling, and signal watchdog functionality have been implemented and continue to evolve. CODAC demonstrated a mechanism by which individual function models coordinate and cooperate with each other to achieve desired higher-level functions. Operation of prototype and real-life experiments verified the application is compliant with the design input and even adaptable and extendable in various system attributes.This paper will introduce the PCS prototyping and its execution as well as integrated operation function in the ITER environment.

Tokamak and ITER plasma with the Quantum Alfven waves

The hydrodynamical model of quantum mechanics based on the Schrodinger equation is combined with the magnetohydrodynamical term to form so called quantum magnetohydrodynamic equation. It is shown that the quantum correction to the Alfven waves follows from this new equation. The possible generalization is considered for the so called nonlinear Schrodinger equation and for the situation where dissipation is described by the Navier-Stokes equation.