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The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design ('Decide & Iterate') methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the 'Decide and Iterate' methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D2O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme hypothesizes that a compact machine offers a route to reduced capital cost that directly tackles the barrier to entry of this potentially transformative technology. History has shown that with an unsolved, complex and highly interdependent design challenge, there is a need to balance exploration of the problem with progress. Almost all complex systems arise from the evolutionary improvement of simpler systems which is an approach the programme has adopted by working through a virtual natural selection of design families towards a single concept consistent with the initiating hypothesis. Issues are uncovered and solved more rapidly this way because the effort is focused on an end. In this current phase, STEP has had to be an agile fast-moving programme to work with what emerges as well as what was planned, to sit with uncertainty and to embrace self-organizing principles. The complex decision-making and compromises in emerging trades have led to a concept respectful of the tight aspect ratio hypothesis which carefully balances cost, performance and deliverability. It remains a high-risk and high-reward programme, but the character of the challenge is better understood building confidence and enhancing capability to advance the evolving design further. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The programme to design plasma scenarios for the Spherical Tokamak for Energy Production (STEP), a reactor concept aiming at net electricity production, seeks to exploit the inherent advantages of the spherical tokamak (ST) while making conservative assumptions about plasma performance. This approach is motivated by the large gap between present-day STs and future burning plasmas based on this concept. It is concluded that plasma exhaust in such a device is most likely to be manageable in a double null (DN) configuration, and that high core performance is favoured by positive triangularity (PT) plasmas with an elevated central safety factor. Based on a full technical and physics assessment of external heating and current drive (CD) systems, it was decided that the external CD is provided most effectively by microwaves. Operation with active resistive wall mode (RWM) stabilization as well as high elongation is needed for the most compact solution. The gap between existing devices and STEP is most pronounced in the area of core transport, owing to high normalized plasma pressure in the latter which changes qualitatively the nature of the turbulence controlling transport. Plugging this gap will require dedicated experiments, particularly on high-performance STs, and the development of reduced models that faithfully represent turbulent transport at high normalized pressure. Plasma scenarios in STEP will also need to be such that edge localized modes (ELMs) either do not occur or are small enough to be compatible with material lifetime limits. The high current needed for a power plant-relevant plasma leads to the unavoidable generation of high runaway electron beam current during a disruption, where novel mitigation techniques may be needed. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) environment will include magnetic, thermal, mechanical and environmental loads far greater than those seen in the Joint European Torus campaigns of the past decade or currently contemplated for ITER. Greater still are the neutron peak dose rates of 10−6 displacements per atom, per second, which in-vessel materials in STEP are anticipated to be exposed to. Reduced activation and high-fluence resilience therefore dominate the materials strategy to support the STEP Programme. The latter covers the full life cycle from downselected compositions and new microstructural developments to irradiation-informed modelling and end-of-life strategies. This article discusses how the materials downselection is oriented in plant power trade-off space, outlines the development of an advanced ferritic-martensitic structural steel, describes the 'Design by Fundamentals' mesoscale modelling approach and reports some of the waste mitigation routes intended to make STEP operations as sustainable as possible. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The design, delivery and operation of a large-scale infrastructure project are challenging at best. For the Spherical Tokamak for Energy Production (STEP) prototype powerplant (SPP), the challenges increased dramatically. In addition to being a large-scale infrastructure project, it is a cutting edge, first of a kind (FOAK) technology demonstrator. The design teams are working in a volatile, uncertain, complex and ambiguous environment, where technology is constantly emerging, maturing and changing. STEP will be unlike any power plant ever built and requires the development of new technologies and capabilities, but also a novel approach to planning and maturing the design. By taking a holistic view of the engineering life cycle from the start, the programme will be better positioned to achieve an SPP that is fit for purpose and can be used to show a path to ultimate commercial viability for subsequent power plants. This paper will review the key challenges in maturing a FOAK fusion power plant and look in depth at how the STEP team are maturing the required capabilities and planning to ensure successful delivery of the SPP. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme is an ambitious but challenging endeavour to design and deliver a prototype fusion power plant. It is a rapid, fast-moving programme, designing a first of a kind device in a Volatile, Uncertain, Complex and Ambiguous (VUCA) environment, and digital tools play a pivotal role in managing and navigating this space. Digital helps manage the complexity and sheer volume of information. Advanced modelling and simulation techniques provide a platform for designers to explore various scenarios and iteratively refine designs, providing insights into the intricate interplay of requirements, constraints and design factors across physics, technology and engineering domains and aiding informed decision-making amidst uncertainties. It also provides a means of building confidence in the new scientific, technological and engineering solutions, given that a full-scale-integrated precursor test is not feasible, almost by definition. The digital strategy for STEP is built around a vision of a digital twin of the whole plant. This will evolve from the current digital shadow formed by system architecting codes and complex workflows and will be underpinned by developing capabilities in plasma, materials and engineering simulation, data management, advanced control, industrial cybersecurity, regulation, digital technologies and related digital disciplines. These capabilities will help address the key challenges of managing the complexity and quantity of information, improving the reliability and robustness of the current digital shadow and developing an understanding of its validity and performance. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) requires high-field magnet designs and has therefore adopted the REBCO-based high-temperature superconductor (HTS) as its current carrier. The HTS enables the toroidal field (TF) coils to be remountable, which unlocks STEP's vertical maintenance approach; however, remountable joints, approximately 18 GJ of stored energy and limited space down the centre of a spherical tokamak, make the TF coils the most challenging. STEP has pursued a passive approach to TF coil quench protection in order to limit coil terminal voltage. Initial results suggest that a solution may rely on tuning internal coil resistance coupled with actively powered heaters. The pre-conceptual inter-coil structure demonstrates acceptable stresses and deflections under steady-state operating conditions and preliminary fault scenarios, and loads are distributed to limit the tensile force on the TF centre rod. Finally, the HTS must operate reliably in a high radiation environment and endure high neutron fluences, ensuring commercially relevant magnet lifetimes. Initial experiments indicate that instantaneous gamma irradiation of HTS has no negative impact on current carrying capacity. Experimental programmes are underway to cold irradiate HTS to fusion-relevant fluences and to develop a method of assuring tape irradiation tolerance using oxygen ions as an analogue for neutrons. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The UK's fusion energy approach has developed over the past 5 years to include government policy initiatives and a range of public sector investments designed to be delivered in partnership with the private sector. These have aimed to create an environment that stimulates innovation and investment to deliver economic as well as scientific and environmental benefits throughout the lifetime of the public sector fusion energy programme. The Spherical Tokamak for Energy Production acts as a focus and anchor for both public and private sector efforts to develop fusion energy, developing the supply chain and potential for Intellectual Property development and export opportunities well ahead of the anticipated STEP completion date of 2040. This is maximized by the UK's approach to a holistic research and innovation programme backed up by a regulatory and skills programme. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

Success of the UK's Spherical Tokamak for Energy Production (STEP) programme requires a robust plasma control system. This system has to guide the plasma from initiation to the burning phase, maintain it there, produce the desired fusion power for the desired duration and then terminate the plasma safely. This has to be done in a challenging environment with limited sensors and without overloading plasma-facing components. The plasma parameters and the operational regime in the STEP prototype will be very different from tokamaks, which are presently in operation. During fusion burn, the plasma regime in STEP will be self-organizing, adding further complications to the plasma control system design. This article describes the work to date on the design of individual controllers for plasma shape and position, magneto hydrodynamic instabilities, heat load and fusion power. Having studied 'normal' operation, the article discusses the philosophy of how the system will handle exceptions, when things do not go exactly as planned. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) prototype powerplant (SPP) will be a first-of-a-kind powerplant—its prime objective is to export electrical power, to the national power transmission system ('grid'), above 100 MWe. As part of a wider issue, addressing the STEP concept design, this article seeks to explore how electrical power will be generated from a spherical tokamak heat source. Accordingly, the following key functions of the SPP power infrastructure are reviewed. Cooling the tokamak: cooling the tokamak while extracting useful thermal energy. Generating power: conversion of thermal energy to electrical energy (power generation). Managing energy: management of the site-wide distribution, storage and energy export. In each of these areas, the design scope, challenges and solution spaces have been discussed. This has shaped the design of the SPP power infrastructure, which in turn has ensured a powerplant design focused on operability and performance. Furthermore, it has been demonstrated that the SPP will achieve its prime objective in generating net power, which is enabled by a unique power infrastructure. Confidence in the ability to generate net power will be refined as the design matures. Finally, this article recommends key opportunities that STEP could use to improve power generation and reduce the parasitic load of the SPP. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

This article describes why Spherical Tokamak for Energy Production (STEP) has been launched, what it aims to achieve (benefits) and, principally, how the whole programme will be delivered (strategy). The article draws on the work of major project delivery and organization design (OD) and applies this to the context of STEP, which is dominated by significant uncertainty in all dimensions (technical, financial, commercial and programmatic), where there is embryonic delivery capability, but where there are also global-scale opportunities. This leads to an approach based on securing and organizing the correct capability from both public and private sectors to work in a collaborative arrangement with a single purpose and, critically, in an operating model designed to manage uncertainty and emerging risks and to exploit opportunities. Placing adaptability at the core of the OD, particularly the ability to deliver emergent strategy through guided empowerment in pursuit of an ambitious aim, is a further development beyond much of the current thinking in major projects. The article concludes with an appendix that translates that programme approach into principles for managing the engineering design work. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The development of nuclear fusion as a safe and virtually limitless power source is receiving growing attention in the context of looming energy crisis and climate change. ITER project stands as the flagship international initiative and is advancing steadily. The construction of the Tokamak Complex is nearly finished, and the assembly of core components has begun on site. Simultaneously, the design is being finalized, and the safety case is becoming more concrete. Current approaches to radiation safety demonstration using 3D nuclear analysis with the Monte Carlo code MCNP require sophisticated artifacts to sew together simulations in separate models for the Tokamak and the rest of the facility. This results in cumbersome studies and, consequently, challengeable conclusions. To address this issue, we have built the an integral MCNP model of the ITER facility: the ITER full model. Along with improvements to the D1SUNED code, we illustrate its computational practicality and pertinence in two meaningful simulations for ITER safety case. This work represents the culmination of a two-decade-long effort of ITER modelling aiming to demonstrate adequate radiation safety. Beyond supporting the remaining design tasks, this model simplifies the corresponding 3D nuclear analysis and improves the robustness of the ITER safety case. In this work, authors present an integral MCNP model of the ITER facility for radiation transport, addressing limitations of separate models for the Tokamak and Tokamak Complex. This model streamlines safety simulations, enhancing the robustness and simplicity of radiation safety analysis. [ABSTRACT FROM AUTHOR]

Quasilinear treatments are widely used for tokamaks to evaluate radio frequency (rf) heating and current drive. Even though the core of a tokamak plasma is weakly collisional, the solution of the linearized kinetic equation is evaluated using unperturbed collisionless trajectories while often treating successive poloidal circuits of the passing (and trapped) particles as uncorrelated or nearly so. In addition, the most important effect of tokamak geometry, the mirror force, is usually mistreated or ignored when obtaining the solution. These concerning aspects of rf treatments are clarified by considering lower hybrid heating and current drive to illustrate that the electrons in resonance with the applied rf are enclosed by narrow collisional boundary layers, and that tokamak geometry makes it necessary to retain poloidal variation when solving a weakly collisional linearized kinetic equation. Other aspects such as collisional boundary layers at the trapped–passing boundary, cyclotron resonances, and the limitations of quasilinear theory are also considered. The new insights lead to a fundamentally different formulation and interpretation of the solution of the linearized Fokker–Planck equation used for rf quasilinear theory in a tokamak, while retaining many of the features that have contributed to its successful application to rf heating and current drive. [ABSTRACT FROM AUTHOR]

The article considers different methods of divertor plasma diagnostics planned for using in the T‑15MD tokamak. Technical problems arising during operation of optical systems in the divertor zone are discussed, including degradation of in-vessel optical elements. The main attention is paid to the conceptual design of the "Passive Spectroscopy in Divertor" diagnostic system. The optical system, including in-vessel mirrors, and methods for protecting its components from the negative effects of plasma are described in detail. Using synthetic diagnostics and numerical simulation methods, the possibility of solving the problem of tomographic reconstruction of the two-dimensional profile of plasma radiation in the T-15MD divertor is demonstrated. Based on the results presented, it was concluded that passive spectroscopy can be used for obtaining data on plasma parameters in the divertor with good spatial resolution, which will make it possible to study the physics of processes and monitor the operation of the T-15MD divertor, including the operation in detachment regime. [ABSTRACT FROM AUTHOR]

Hydrogen isotope mixing phenomena in tokamak plasmas are analyzed using global full-f gyrokinetic simulations. Model plasma parameters are chosen based on hydrogen isotope pellet experiments on JET, in which fast hydrogen isotope mixing in the timescale of the energy confinement time τ E occurred after injecting deuterium (D) pellets into hydrogen (H) plasmas. Two numerical experiments are conducted using plasma profiles before and after the pellet injection. In both cases, turbulent fluctuations in the plasma core are characterized by ion temperature gradient driven turbulence. In the former case, the density profile of bulk H ions is kept in a quasi-steady state, and the particle confinement time of bulk H ions τ H is an order of magnitude longer than τ E . In the latter case, the density profiles of bulk H ions and pellet D ions show transient relaxation in the timescale of τ E , and both τ H and the timescale of particle pinch τ D become comparable to τ E , indicating the fast hydrogen isotope mixing. In the toroidal angular momentum balance, it is found that the enhanced ion particle transport leading to the fast hydrogen isotope mixing is driven by the toroidal field stress. [ABSTRACT FROM AUTHOR]

Tokamak disruptions are associated with breaking magnetic surfaces, which makes magnetic field lines chaotic in large regions of the plasma. The enforcement of quasi-neutrality in a region of chaotic field lines requires an electric potential that has both short and long correlation distances across the magnetic field lines. The short correlation distances produce a Bohm-like diffusion coefficient ∼ T e / e B and the long correlation distances a T produce a large scale flow ∼ T e / e B a T . This cross-field diffusion and flow are important for sweeping impurities into the core of a disrupting tokamak. The analysis separates the electric field in a plasma into the sum of a divergence-free, E → B , and a curl-free, E → q , part, a Helmholtz decomposition. The divergence-free part of E → determines the evolution of the magnetic field. The curl-free part enforces quasi-neutrality, E → q = − ∇ → Φ q . Magnetic helicity evolution gives the required boundary condition for a unique Helmholtz decomposition and an unfortunate constraint on steady-state tokamak maintenance. [ABSTRACT FROM AUTHOR]

Gyrokinetic simulation of the toroidal rotation of plasma with a stochastic magnetic field perturbation is carried out. The simulation results suggest that the stochastic magnetic perturbation drives the plasma to toroidally rotate through the ambipolar radial electric field E r established on the timescale of electron transit time. It is found that this spontaneous flow driven on the timescale less than an ion–ion collision time is the parallel return flow of the E r × B drift. The ion–ion collisional effect further changes the parallel return flow to the toroidal rigid-body flow after a few ion–ion collision times. This simulation result is consistent with the toroidal rigid-body rotation observed in tokamak experiments with a stochastic layer induced by the externally applied resonant magnetic perturbation. [ABSTRACT FROM AUTHOR]

The tearing mode stability of an inverse aspect-ratio expanded tokamak plasma equilibrium of general shape is investigated using asymptotic matching techniques. Particular emphasis is placed on the conservation of toroidal electromagnetic angular momentum. The TJ code, which is a specific implementation of the results of the investigation, is described. [ABSTRACT FROM AUTHOR]

Neoclassical transport in a non-isothermal plasma in which each plasma species has different equilibrium temperature is investigated by solving the drift kinetic equation with a Fokker–Planck (FP) collision operator in a circular tokamak model. Since it is known that a linearized FP operator does not have a self-adjoint property in a non-isothermal plasma, approximate models are developed for comparison to intend to have the self-adjoint property in the non-isothermal case. To achieve this, we set a common temperature that the system should reach after a long time, and the individual temperature of each particle species is expressed by a parameter to measure a shift of the individual temperature from the common one. Then, both the Vlasov part and the collision term of the kinetic equation are expanded around the common temperature, taking the temperature shift parameter up to the first order. It is found that the lowest order collision term of expansion preserves the self-adjointness while the first-order, nonlinear FP term does not. A large difference of the ion heat neoclassical transport is found in comparison between the developed models with and without the self-adjointness and the original FP collision term in the non-isothermal plasma, especially in a strong temperature equilibration regime, showing that a contribution of the collision term without the self-adjointness seems to be significant. Furthermore, when an impurity species is included, the result is complicated where the usual enhancement in the main ion particle transport coefficient, due to the impurity effect, is rather suppressed with the increase in the ion heat transport coefficients by the non-isothermal effects. [ABSTRACT FROM AUTHOR]

Helicon waves are thought to be promising in various tokamaks, such as DIII-D, because they can penetrate reactor-grade high-density cores and drive the off-axis current with higher efficiency. In the frequency regime ∼ 476 MHz, both slow electrostatic and fast electromagnetic helicon waves can coexist in DIII-D. If the antenna parasitically excites the slow mode, these waves can propagate along the magnetic field line into the scrape-off layer (SOL). Although the importance of the misalignment of the Faraday screen and the electron density in the SOL on the excitation and propagation of slow modes is well known, the conditions for minimizing slow mode excitation have yet to be optimized. Using the Petra-M simulation code in the 2D domain, we analyze the effects of the misalignment of the antenna in the poloidal direction, the misalignment of the Faraday screen in the toroidal direction, and the density in front of the antenna on slow mode generation. Our results suggest that the misalignment of the Faraday screen is a critical factor in reducing the slow mode and that the misalignment angle should be below ∼ 5° to minimize the slow wave excitation. When the electron density is higher than 3.5 × 10 18 m − 3 in the SOL, the generation of the slow mode from the antenna is minimized and unaffected by the misalignment of the Faraday screen. [ABSTRACT FROM AUTHOR]

The SPARC tokamak is a high-field, Bt0 ∼12 T, medium-sized, R0 = 1.85 m, tokamak that is presently under construction in Devens, MA, led by Commonwealth Fusion Systems. It will be used to de-risk the high-field tokamak path to a fusion power plant and demonstrate the commercial viability of fusion energy. SPARC's first campaign plan is to achieve Qfus > 1 using an ICRF-heated, <10 MW, high current, Ip ∼ 8.5 MA, L-mode fueled by D–T gas injection, and its second campaign will investigate H-mode operations in D–D. To facilitate plasma control and scientific learning, a targeted set of ∼50 plasma diagnostics are being designed and built for operation during these campaigns. While nearly all diagnostics are based on established techniques, the pace of deployment, relative to the first plasma, and the harshness of the thermal, electromagnetic, and radiation environment are unprecedented for medium-sized tokamaks. An overview of the SPARC diagnostic set is given, providing context to further details communicated by the SPARC team in companion publications that are system-specific. The system engineering philosophy for SPARC diagnostics is outlined, and the design and engineering verification process for components inside and outside the primary vacuum boundary are described. Diagnostics are mounted directly to the vacuum vessel as well as housed within a series of eight midplane and 24 off-midplane replaceable port plugs. With limited exceptions, signal conditioning, digitization electronics and cameras as well as lasers and microwave sources are localized to a series of five Diagnostic Lab spaces, totaling ∼350 m2, located >15 m from the center of the tokamak, on the other side of a 2.4 m concrete shielding wall. A series of 31 large-scale penetrations have been included in the SPARC Tokamak Hall to facilitate integration of early campaign diagnostics and to provide upgradability. [ABSTRACT FROM AUTHOR]

Neutron measurement is the primary tool in the SPARC tokamak for fusion power (Pfus) monitoring, research on the physics of burning plasmas, validation of the neutronics simulation workflows, and providing feedback for machine protection. A demanding target uncertainty (10% for Pfus) and coverage of a wide dynamic range (>8 orders of magnitude going up to 5 × 1019 n/s), coupled with a fast-track timeline for design and deployment, make the development of the SPARC neutron diagnostics challenging. Four subsystems are under design that exploit the high flux of direct DT and DD plasma neutrons emanating from a shielded opening in a midplane diagnostic port. The systems comprise a set of ∼15 flux monitors, mainly ionization chambers and proportional counters for measurement of the neutron yield rate, two independent foil activation systems for measurement of the neutron fluence, a spectrometric radial neutron camera for poloidal profiling of the plasma emissivity, and a high-resolution magnetic proton recoil spectrometer for measurement of the core neutron spectrum. Together, the four systems ensure redundancy of sensors and methods and aim to provide high resolutions of time (10 ms), space (∼7 cm), and energy (<2% at 14 MeV). This paper presents the broader objectives behind the preliminary design of the SPARC neutron diagnostics and discusses the ongoing studies on neutronics, detector comparisons, prototyping, and integration with the unique infrastructure of SPARC. Engineering details of the four subsystems and the concepts for in situ neutron calibration are also highlighted. [ABSTRACT FROM AUTHOR]

The fast-ion phase-space distribution function in the magnetic fusion devices is always underdiagnosed, and every new fast-ion diagnostic should be carefully assessed before installation to minimize redundancies in measurements and maximize the information from the yet undiagnosed part of the fast-ion phase space distribution function. Here, we present a novel method of assessing the added value of a considered fast-ion diagnostic, taking actual geometry and an existing set of fast-ion diagnostics into account. The new method is based on a reformulation of the diagnostic weight functions in constants of motion (COM). We compare the proposed method with the previous approach using Monte Carlo simulations. [ABSTRACT FROM AUTHOR]

The European DEMO (EU-DEMO) reactor studies within EUROfusion aim to develop a fusion power plant concept. The large tokamak device needs an auxiliary heating power which, at the present stage, is provided by the Electron Cyclotron (EC) heating system with up to 130 MW foreseen to reach different regions of plasma for heating, suppression of instabilities and the possibility to support ramp-up and ramp-down phases. The present conceptual design of the system is based on 2 MW coaxial-cavity gyrotron sources, a transmission line (TL) using both circular corrugated waveguides and quasi-optical evacuated multi-beam TLs, and mirror antennas located in the Equatorial Port. In order to create a modular system, the sources are grouped in 'clusters', whose powers are combined in the quasi-optical TL, up to the tokamak building, where they are split and routed as single waveguides. In the launcher, they are combined together again on the launching mirrors, to save space for the apertures in the Breeding Blanket. The present EC heating system has a certain flexibility to adapt to changing design guidelines. The development status of the system is presented. [ABSTRACT FROM AUTHOR]

The work reported in this paper addresses two aspects. In the first part, numerical simulations are conducted to examine the impact of magnetic equilibrium shaping (elongation and triangularity), on both conventional Ion Temperature Gradient (ITG) modes and Short Wavelength ITG modes. This analysis is performed considering the experimental profiles and parameters of the ADITYA-U tokamak, employing the nonlinear global gyrokinetic Particle-in-Cell code ORB5. The linear and nonlinear collisionless electrostatic simulation of these modes are carried out with kinetic ions and adiabatic electrons. From the linear results, it has been observed that the magnetic equilibrium shaping slighty reduced the growth rates and widened the spectrum, and the maxima of growth rate curve is shifted to higher toroidal wave number. We find that the heat flux is reduced by a significant ≃ 35 % for the true circular magnetohydrodynamic magnetic equilibrium as compared to ad hoc concentric circular equilibrium reported in Singh et al (2023 Nucl. Fusion 63 086029). A further ≃ 10 % reduction in the heat flux is seen with magnetic equilibrium shaping. In the second part, linear collisionless electrostatic simulation studies of ITG coupled with fully kinetic electrons (both trapped and passing electrons are treated kinetically) using a drift-kinetic ordering is performed. It can be seen from the linear results that, in presence of kinetic electrons, the growth rate and real frequency of the ITG mode are increased significantly for ADITYA-U parameters and a mode propagating in the electron diamagnetic direction is identified at high toroidal wavenumbers. [ABSTRACT FROM AUTHOR]

Introduction: This work explores the use of eXplainable artificial intelligence (XAI) to analyze a convolutional neural network (CNN) trained for disruption prediction in tokamak devices and fed with inputs composed of different physical quantities. Methods: This work focuses on a reduced dataset containing disruptions that follow patterns which are distinguishable based on their impact on the electron temperature profile. Our objective is to demonstrate that the CNN, without explicit training for these specific mechanisms, has implicitly learned to differentiate between these two disruption paths. With this purpose, two XAI algorithms have been implemented: occlusion and saliency maps. Results: The main outcome of this paper comes from the temperature profile analysis, which evaluates whether the CNN prioritizes the outer and inner regions. Discussion: The result of this investigation reveals a consistent shift in the CNN's output sensitivity depending on whether the inner or outer part of the temperature profile is perturbed, reflecting the underlying physical phenomena occurring in the plasma. [ABSTRACT FROM AUTHOR]

Time delay and velocity estimation methods have been widely studied subjects in the context of signal processing, with applications in many different fields of physics. The velocity of waves or coherent fluctuation structures is commonly estimated as the distance between two measurement points divided by the time lag that maximizes the cross correlation function between the measured signals, but this is demonstrated to result in erroneous estimates for two spatial dimensions. We present an improved method to accurately estimate both components of the velocity vector, relying on three non-aligned measurement points. We introduce a stochastic process describing the fluctuations as a superposition of uncorrelated pulses moving in two dimensions. Using this model, we show that the three-point velocity estimation method, using time delays calculated through cross correlations, yields the exact velocity components when all pulses have the same velocity. The two- and three-point methods are tested on synthetic data generated from realizations of such processes for which the underlying velocity components are known. The results reveal the superiority of the three-point technique. Finally, we demonstrate the applicability of the velocity estimation on gas puff imaging data of strongly intermittent plasma fluctuations due to the radial motion of coherent, blob-like structures at the boundary of the Alcator C-Mod tokamak. [ABSTRACT FROM AUTHOR]

Disruptions in tokamak nuclear reactors, where plasma confinement is suddenly lost, pose a serious threat to the reactor and its components. Classifying discharges as disruptive or non-disruptive is crucial for effective plasma operation and advanced prediction. Traditional disruption identification systems often struggle with noise, variability, and limited adaptability. To address these challenges, we propose an enhanced stacking generalization model called the "Double-Phase Stacking Technique" integrated with Pool-based Active Learning (DPST-PAL) for designing a robust classifier with minimal labor cost. This innovative approach improves classification accuracy and reliability using advanced data analysis techniques. We trained the DPST-PAL model on 162 diagnostic shots from the Aditya dataset, achieving a high accuracy of 98% and an F1-score of 0.99, surpassing conventional methods. Subsequently, the deep 1D convolutional predictor model is implemented and trained using the classified shots obtained from the DPST-PAL model to validate the reliability of the dataset, which is tested on 47 distinct shots. This model accurately predicts the disruptions 7–13 ms in advance with 93.6% accuracy and exhibited no premature alarms or misclassifications for our experimental shots. [ABSTRACT FROM AUTHOR]

Here, we present the design and first calibration results of a new single-channel Fast-Ion D-Alpha (FIDA) spectrometer to be employed at the National Spherical Torus Experiment Upgrade (NSTX-U). The Czerny–Turner-type spectrometer uses a custom-designed aspherical lens setup instead of mirrors and achieves excellent spectral resolution, with high photon throughput through a round-to-linear fiber bundle, and camera frame rates around 8.4 kHz. The spectrometer uses a blocking bar to avoid saturation effects of the cold D-alpha emission line and will allow for detailed studies of the fast-ion confinement in NSTX-U. Expected synthetic spectra predicted with the TRANSP and FIDASIM codes show that the spectral range from 648.5 to 658 nm will sufficiently cover halo, the red-shifted beam emission, and the blue-shifted portion of FIDA emission in NSTX-U, which is sufficient for fast-ion transport studies of co-rotating fast ions. [ABSTRACT FROM AUTHOR]

A multi-energy soft x-ray diagnostic is planned to operate in the small aspect ratio tokamak (SMART), consisting of five cameras: one for core measurements, two for edge, and two for divertors. Each camera is equipped with four absolute extreme ultra-violet diodes, with three of them filtered by Ti and Al foils for C and O line emissions, respectively, and Be foils for temperature measurements. In addition, two spectrometers will be installed with a vertical line of sight for impurity control. This study introduces a synthetic model designed to characterize radiated power and soft x-ray emissions. The developed code extracts the radiated power and Zeff values by leveraging distributions of electron density, temperatures, and impurity concentrations. The investigation is centered on the predicted scenarios of SMART's first phase of operation (Ip = 100 kA; Bt = 0.1 T), employing a double-null configuration with positive and negative triangularity. The anticipated impurities encompass C (1%) and Fe (0.01%) from the vessel, as well as O and N (0.1%) from air and water. For simplicity, the distribution is assumed to be homogeneous within the plasma, considering different mixtures with Zeff values ranging between 1 and 2. Finally, the model estimates signal strength for the diagnostic design, proving its feasibility. [ABSTRACT FROM AUTHOR]

Near ion-cyclotron frequency (fci) fluctuations, such as those originating from Global and Compressional Alfvén Eigenmodes (GAEs/CAEs), are expected to be present in future fusion reactors but are not well understood due to the limited availability of core measurements in present-day tokamaks. The measurement bandwidth of the Radial Interferometer–Polarimeter (RIP) diagnostic has been upgraded from 1 to 5 MHz to detect these fluctuations in DIII-D. RIP adopts the three-wave technique for simultaneous polarimetric and interferometric measurements. Solid-state microwave sources operating at 650 GHz are used as probe beams and provide 5 MHz bandwidth for both polarimetric and interferometric measurements. Bandwidths of related hardware, including mixer amplifier, signal cable, and digital phase demodulator, are increased correspondingly. Measurement noise is minimized by reducing the time delay between reference and probe signals to nanosecond level and employing correlation-based techniques. Using the upgraded diagnostic, CAE/GAE-like bursting fluctuations are observed in neutral-beam heated plasmas with toroidal magnetic field Bφ ≈ 1 T. Current upgrades being undertaken would enable the evaluation of toroidal mode number for these modes. This work opens the possibility of better understanding near ion-cyclotron frequency fluctuations in fusion relevant plasmas. [ABSTRACT FROM AUTHOR]

The in-vessel silicon diode arrays that are used for soft x-ray detection in many tokamaks are sensitive to neutron damage, making them unsuitable for burning plasma devices such as SPARC. In such a device, the silicon diodes would need to be placed far from the plasma—limiting their field of view—or an alternative detector could be used. Here, we present the design of a camera containing an array of chemical vapor deposition single-crystal diamonds, which will be placed in the upper and lower port plugs of the SPARC tokamak with a large enough view of the poloidal cross section to enable tomographic inversion. The camera design presented here is optimized to provide a wide field of view of the poloidal cross section. Simulated plasma conditions are used to estimate the x-ray signal that this detector array will receive and to fine-tune the camera placement within the tokamak. [ABSTRACT FROM AUTHOR]

In this paper, we report the excitation of coherent density and potential fluctuations induced by magnetohydrodynamic (MHD) activity in the edge plasma region of ADITYA-U tokamak. When the amplitude of the MHD mode, mainly the m/n = 2/1, increases beyond a threshold value, | B ̃ θ | / B θ ∼ 0.3 % – 0.4 % , coherent oscillations in the density and potential fluctuations are observed having the same frequency as that of the MHD mode. The mode numbers of these MHD induced density and potential fluctuations are obtained by Langmuir probes placed at different radial, poloidal, and toroidal locations in the edge plasma region. Detailed analyses of these Langmuir probe measurements reveal that the coherent mode in edge potential fluctuation has a mode structure of m/n = 2/1, whereas the edge density fluctuation has an m/n = 1/1 structure. It is further observed that beyond the threshold, the coupled power fraction scales almost linearly with the magnitude of B ̃ θ / B θ fluctuations. Furthermore, the rise rates of the coupled power fraction for coherent modes in density and potential fluctuations are also found to be dependent on the growth rate of magnetic fluctuations. The disparate mode structures of the excited modes in density and plasma potential fluctuations suggest that the underlying mechanism for their existence is the coupling of even harmonics of potential to the odd harmonics of pressure due to 1/R dependence of the toroidal magnetic field. [ABSTRACT FROM AUTHOR]

Velocity, density, and magnetic fluctuations of the geodesic acoustic mode (GAM) have been measured using the Doppler backscattering system, Faraday-effect polarimeter-interferometer, and external pick-up coils in the Experimental Advanced Superconducting Tokamak. Simultaneous measurements of density and velocity fluctuations at the midplane and top of plasmas demonstrate that m = 1 density fluctuations are quantitatively balanced by the compression of perpendicular flow fluctuations. Furthermore, internal magnetic fluctuations associated with GAM have now been directly measured by laser-based Faraday-effect polarimetry for the first time. Line-averaged magnetic fluctuations (up to 16 Gauss, B ̃ ¯ R , GAM B T ∼ 0.066 %) are significantly larger than those extrapolated from edge coils (a few Gauss) and that magnetic fluctuations increase with β. The observed discrepancy between finite β theory and experimental data indicates the need for further theoretical investigations. [ABSTRACT FROM AUTHOR]

In this work, we explore the triangularity effects on turbulent transport employing global gyrokinetic simulations performed with the ORB5 code. Numerous experiments on the Tokamak á Configuration Variable (TCV) and, more recently, on the DIII-D machine, have demonstrated superior confinement properties in L-mode of negative triangularity (NT) over positive triangularity (PT) configuration. This presents a particularly attractive scenario, as L-mode operation eliminates or significantly mitigates the presence of hazardous edge-localized modes (ELMs). However, a full theoretical understanding of all these observations remains elusive. Specifically, questions remain about how NT improvements can extend to the core where triangularity is very small, and whether these improvements can scale to larger devices. This paper addresses these two questions. Our analysis is divided into two parts: we first demonstrate that the confinement enhancement in NT configurations arises from the interdependent edge-core dynamics, and then we present the results of a system size scan. Crucially, we find that the relative turbulent transport reduction of NT over PT appears not to be contingent on machine dimensions or fluctuation scales and is moreover robust with respect to variations in plasma profiles. This insight underscores the fundamental nature of the NT confinement advantage and paves the way for its potential application in future fusion devices, regardless of their size. [ABSTRACT FROM AUTHOR]

In its first phase of exploitation, JT-60SA will be equipped with an inertially cooled divertor, which can sustain heat loads of 10 MW/m[sup 2] on the targets for a few seconds, which is much shorter than the intended discharge duration. Therefore, in order to maximize the duration of discharges, it is crucial to develop operational scenarios with a high radiated fraction in the plasma edge region without unacceptably compromising the scenario performance. In this study, the core and edge conditions of unseeded and neon-seeded deuterium H-mode scenarios in JT-60SA were investigated using METIS and SOLEDGE3X-EIRENE codes. The aim was to determine whether, and under which operational conditions, it would be possible to achieve heat loads at the targets significantly lower than 10 MW/m[sup 2] and potentially establish a divertordetached regime while keeping favorable plasma core conditions. In first analysis, an investigation of the edge parameter space of unseeded scenarios was carried out. Simulations at an intermediate edge power of 15 MW indicate that, without seeded impurities, the heat loads at the targets are higher than 10 MW/m[sup 2] in attached cases, and achieving detachment is challenging, requiring upstream electron densities at least above 4 χ 10[sup 19] m[sup -3]. This points toward the need for impurity injection during the first period of exploitation of the machine. Therefore, neon seeding simulations were carried out, performing a seeding rate scan and an injected power scan while keeping the upstream electron density at the separatrix at 3 χ 10[sup 19] m[sup -3]. They show that at 15 MW of power injected into the edge plasma, the inner target is easily detached and presents low heat loads when neon is injected. However, at the outer target, the heat fluxes are not lowered below 10 MW/m[sup 2], even when the power losses in the edge plasma are equal to 50% of the power crossing the separatrix. Therefore, the tokamak will probably need to be operated in a deep detached regime in its first phase of exploitation for discharges longer than a few seconds. In the framework of core-edge integrated modeling, using METIS, the power radiated in the core was computed for the most interesting cases. [ABSTRACT FROM AUTHOR]

Microinstabilities drive turbulent fluctuations in inhomogeneous, magnetised plasmas. In the context of magnetic confinement fusion devices, this leads to an enhanced transport of particles, momentum and energy, thereby degrading confinement. In this work, we describe an application of the adjoint method to efficiently determine variations of gyrokinetic linear growth rates on a general set of external parameters in the local $\delta f$ -gyrokinetic model. We then offer numerical verification of this approach. When coupled with gradient-based techniques, this methodology can facilitate the optimisation process for the microstability of the confined plasmas across a high-dimensional parameter space. We present a numerical demonstration wherein the ion-temperature-gradient instability growth rate in a tokamak plasma is minimised with respect to flux surface shaping parameters. The adjoint method approach demonstrates a significant computational speed-up compared with a finite-difference gradient calculation. [ABSTRACT FROM AUTHOR]

This paper presents a first implementation of gradient, divergence, and particle tracing schemes for the EMC3 code, a stochastic 3D plasma fluid code widely employed for edge plasma and impurity transport modeling in tokamaks and stellarators. These schemes are essential to accommodate plasma drift computations, which are currently absent in the code. Plasma drifts have been recognized to significantly influence transport of particles and energy, and their inclusion in future code upgrades will substantially enhance the code's predictive capabilities. For gradient and divergence calculations, we introduce a second‐order least‐squares gradient scheme. We confirm the second‐order convergence properties and assess the accuracy of several analytical test cases in the presence of synthetic noise. In the second part of this paper, we employ the validated gradient scheme in a fourth‐order Runge–Kutta particle tracing scheme to trace a particle through a generic drift velocity field. The impact of synthetic noise on the scheme's performance is investigated by evaluating various error metrics. We find that the implemented schemes function as intended and exhibit sufficient accuracy to enable drift computations. [ABSTRACT FROM AUTHOR]

A new integrating model has been developed to couple tokamak edge multiscale magnetohydrodynamic (MHD) events and transport simulations, such as edge‐localized mode (ELM) cycles. As a proof of principle, we first start from a set of three‐field two‐fluid model equations, which includes the pressure, current, and vorticity. The equations are separated into the slowly evolving part of the axisymmetric component by taking a time average of the axisymmetric component. The time‐averaged fluxes, which are quadratic in fluctuating quantities, act as driven terms for the time‐averaged axisymmetric quantities that determine the plasma transport, and therefore the large‐scale evolution of the plasma profiles. Then the HL‐2A's ELM cycles are simulated using the model. Good agreements of ELM size and pedestal recovery time have been achieved for the solutions obtained from the coupled simulation compared with experiment. For one ELM cycle simulation, the coupled code can achieve a speedup of a factor of up to 30 over standalone code. [ABSTRACT FROM AUTHOR]

The magnetic topology is a critical issue in fusion plasma research. An example is the Resonant Magnetic Perturbation (RMP), which controls the edge transport in tokamaks. However, the physics of how the RMP affects edge transport is not clear. One reason is the transport process on the stochastic magnetic field is poorly understood. This study examines anisotropic heat diffusion numerically to understand heat transport in stochastic magnetic fields. We developed a numerical model of an anisotropic temperature diffusion model, where the significant deviation of the parallel and perpendicular thermal conductivity exists. We applied this implementation to the realistic stellarator geometry with the stochastic magnetic field in the edge. The smooth temperature profile is obtained for the large perpendicular diffusion, although the magnetic field is stochastic. However, for another case of significant parallel diffusion, the small flattening of the temperature on the magnetic island in the stochastic region appears. That result suggests that the stochastic magnetic field can keep the finite temperature gradient if the connection length of the magnetic field line in the stochastic region is sufficiently long. [ABSTRACT FROM AUTHOR]

The effect of differential biasing of the divertor plates in a tokamak is analyzed using SOLPS‐ITER modeling with account of drifts and currents in the SOL. The ASDEX‐Upgrade like geometry and transport parameters adjusted for modeling of the semi‐detached nitrogen‐seeded plasma of shot #28903 are used in analysis. The additional voltage is applied to the outer divertor plate with respect to the grounded vacuum chamber and the inner divertor plate. When high negative voltage is applied, the high field side high density region almost disappears and the divertor plates become more symmetrical than without biasing, while in the case with positive voltage, the formation of the X‐point radiator begins, and both divertor plates become more detached. The results are qualitatively consistent with the experiments conducted on the JFT‐2 M and DIII‐D tokamaks. In addition, both signs of the applied voltage lead to a significant redistribution of the nitrogen within the SOL. Therefore, the possibility of edge plasma control by biasing is demonstrated. [ABSTRACT FROM AUTHOR]

Edge localized modes resulted from magnetohydrodynamic (MHD) instabilities in the pedestal region are a significant concern for future tokamaks. In this work, Ci‐Liu‐Ti (CLT), an MHD code in the three‐dimensional toroidal geometry, is applied for the linear simulation of the ideal pedestal MHD instabilities. The simulations are performed for the experimental advanced superconducting tokamak‐like elongated divertor configuration with large triangularity, which is generated by the high‐accuracy free‐boundary equilibrium solver (CLT‐EQuilibrium, i.e., CLT‐EQ) developed recently. The present work focuses on the influence of the magnetic shear, which is scanned by adjusting the pedestal current with a fixed pedestal pressure profile. As the pedestal current increases, both the local (Slocal) and global (Sglobal) magnetic shear decrease. The ballooning mode is destabilized along with the decrease of Slocal, and stabilized when Slocal is negative for the whole region of bad curvature, which implies the access of the second stable region. Further increase of the pedestal current leads to the destabilization of the kink mode, which is stabilized again until Sglobal is negative at the location of significant gradient of current density. The simulated results are consistent with the findings in Radovanovic et al. Nucl. Fusion 62 (2022) 086004 and C. K. Sun et al. Phys. Plasmas 25 (2018) 082106, which indicates the availability of CLT in the linear simulation of the ideal pedestal instabilities. [ABSTRACT FROM AUTHOR]

A diamagnetic diagnostic system is designed, prepared and installed at Aditya Upgrade Tokamak. Also, the raw data is acquired in a specified acquisition system with suitable front-end electronics. The entire system consists of separate loops and is first calibrated in the presence of all coils, which conduct electricity during the plasma discharge. Then, the change in the toroidal magnetic field due to the plasma is measured using the diamagnetic system, and hence, several plasma parameters including poloidal beta, stored energy, etc. are estimated. The measurements were then repeated during various tests of the tokamak plasma, and several interesting facts were observed. The present paper covers the entire concept of designing magnetic diagnostics – from basic considerations to final data interpretation. [ABSTRACT FROM AUTHOR]

Identification of physical properties of materials is very important because they are in general unknown. Furthermore, their direct experimental measurement could be costly and inaccurate. In such a situation, a cheap and efficient alternative is to mathematically formulate an inverse, but difficult, problem that can be solved, in general, numerically; the challenge being that the problem is, in general, nonlinear and ill-posed. In this paper, the reconstruction of a lower-order unknown time-dependent coefficient in a Cahn–Hilliard-type fourth-order equation from an additional integral observation, which has application to characterizing the nonlinear saturation of the collisional trapped-ion mode in a tokamak, is investigated. The local existence and uniqueness of the solution to such inverse problem is established by utilizing the Rothe method. Moreover, the continuous dependence of the unknown coefficient upon the measured data is derived. Next, the Tikhonov regularization method is applied to recover the unknown coefficient from noisy measurements. The stability estimate of the minimizer is derived by investigating an auxiliary linear fourth-order inverse source problem. Henceforth, the variational source condition can be verified. Then the convergence rate is obtained under such source condition. [ABSTRACT FROM AUTHOR]

We present the design and laboratory tests for a new Q-band frequency tunable Doppler backscattering (DBS) system suitable for probing poloidal wavenumber kñ = 6–8 cm−1 density fluctuations and their flow velocities in the pedestal and scape-off layer (SOL) of the DIII-D tokamak. This system will provide new measurements in the increasingly important and under-diagnosed far pedestal and SOL plasma regions. These results are important for experimental transport studies and necessary for the validation of transport models, both of which are important to fusion energy research. The use of a single tunable frequency reduces the complexity and potential failure points as compared to a multichannel system. This new system utilizes a 33–50 GHz tunable source and will be integrated into the current V-band DBS in DIII-D using a broadband Q- and V-band multiplexer. A full-scale mockup of the quasi-optical system was used to test and optimize the performance. These tests include beam profile measurements at different distances (and angles) from a paraboloidal focusing and steering mirror. The measurements cover the full frequency range 33–75 GHz of the integrated/combined Q–V band DBS system and target a large radial coverage of the low-field side of the plasma from ρ = 1.1 to ρ = 0.5, where ρ is the normalized flux surface radial coordinate. [ABSTRACT FROM AUTHOR]

Recent experiments at the ASDEX Upgrade tokamak have provided the first ever measurements from the imaging heavy-ion beam probe. In this work, we show that the developed simulation framework can reproduce qualitatively the measurement's observed shape and position. Quantitatively, we demonstrate that the model reproduces, within the experimental uncertainties, the observed signal levels. A detailed explanation of the synthetic model is presented, along with the calibration of the optical setup that reproduces the measurements. [ABSTRACT FROM AUTHOR]

The Laser Blow-Off (LBO) impurity injection system is a crucial tool for studying impurity transport and plasma behavior. Conducting proactive impurity transport research is challenging on experimental advanced superconducting tokamak (EAST) due to the uncontrollable generation of impurity sources; therefore, it is necessary to develop a laser blow-off impurity injection system for injecting controlled trace impurity particles. This study presents the design and test results of an LBO system for the EAST. The system aims to provide precise and repeatable control over the timing and quantity of impurity injection. The system primarily consists of a laser source, two mirrors, a moveable focusing lens, a target material, and a vacuum system. The movement of the focusing lens is achieved by a three-dimensional displacement system. The operation of the system is completed by a remote control system. With the accurate control system, the laser spot diameter is adjustable, allowing for modification of impurity injection quantity. The test results demonstrate that the system can rapidly detect external trigger signals and ensure precise timing for the impurity injection. Furthermore, this system can also quickly change the focal point of the laser spot, addressing the requirements for impurity injections during the experiments with less than 0.4 mm position error for laser spot focusing. Test results have shown that the aluminum film material can be peeled off by the LBO system when the laser energy exceeds 650 mJ and the smallest ablation spot is about 1 mm. This study is of significant importance for conducting plasma impurity transport research on the EAST. [ABSTRACT FROM AUTHOR]

Following successful campaigns on the compact high field spherical tokamak ST40, resistive gold foil bolometers have now been installed to measure the radiated power profile. Positioned on the midplane, two bolometer cameras offer perspectives of the horizontal and vertical planes, while the third camera, situated above the midplane, provides a vertical view of the top divertor plates, including the X-point and the strike points. These cameras use a commercial off-the-shelf data acquisition platform with modules specifically designed for bolometers in a Wheatstone bridge configuration, providing both phase sensitive detection and in situ calibration. This paper presents an overview of the design and functionality of all three cameras and the commissioning of the horizontal plane camera. The commissioning of the horizontal camera shows data from a range of plasma pulses, presenting a positive correlation between the relative profiles measured from the bolometer camera and an absolute extreme ultraviolet diode camera viewing along similar lines of sight. [ABSTRACT FROM AUTHOR]

First proton production rates from the d(d,p)t reaction in the Mega Amp Spherical Tokamak Upgrade (MAST-U) are measured. The data were taken during the MAST-U experimental campaign with an upgraded version of the proton detector (PD) previously used in MAST. The new detector array consists of three collimated silicon surface barrier detectors with a depletion depth of 300 μm and a collimated 120 μm thick diamond detector, mounted on the MAST-U reciprocating probe arm. This array measures the energies of unconfined energetic 3 MeV protons and 1 MeV tritons mainly produced by beam-thermal DD reactions during neutral beam injection heating. Diamond detectors have the potential to be uniquely suited to detect charged fusion products as they promise to be much more radiation resistant and much less sensitive to temperature variations compared to silicon-based detectors. Using silicon and diamond-based detectors simultaneously allowed us to directly compare the performance of these two detector types. PD particle rates measured during different plasma scenarios are presented and compared to neutron rates measured using the neutron camera upgrade and TRANSP predictions. [ABSTRACT FROM AUTHOR]

The SPARC tokamak will be equipped with a hard X-ray (HXR) monitor system capable of measuring the bremsstrahlung emission from runaway electrons with photon energies in excess of about 100 keV. This diagnostic will detect the formation of runaway electron beams during plasma start-up and inform the plasma control system to terminate the discharge early to protect the machine. In this work, we present a 0D estimate of the HXR emission in SPARC during plasma start-up. Then we discuss the characterization of a prototype of the HXR monitor. The detector mounts a 1 × 1-in.2 LaBr3 inorganic scintillator coupled with a photomultiplier tube and has been tested with γ-ray sources to find its dynamic range. Finally, two possible modes of operation for spectroscopic and current mode measurements on SPARC are proposed. [ABSTRACT FROM AUTHOR]