Step Tokamak Research Articles

Approach to spatial integration on a novel and complex major project – STEP concept Tokamak

The purpose of this work was to establish an approach to drive the development, at pace, of a commercially relevant Spherical Tokamak Concept. Building on the table of parameters and idealised geometry generated from the work of S. Muldrew et al. [1], applying core principles, tools, and approaches. The generation and utilisation of a reflective 3D Spatial Model to represent the current concept, now forms a strong basis for identifying, evaluating, and delivering future integrated decisions, developed within a 16 Month timeframe.STEP is an ambitious programme, tasked with delivering a prototype plant by 2040 as a pathway to commercial fusion energy. Tackling the integration of a system that itself is based on emergent technologies, significant assumptions (e.g. Plasma Physics) and uncertainties is a significant challenge whilst maintaining acceptable levels of commercial relevance and feasibility. This paper shares the developments and approaches to collaboration and communication taken, including the management of a graphical representation (CAD or Computer Aided Design) to overcome this challenge.The work conducted has resulted in a single point CAD (utilising CATIA software) model representing the current integration status including an appropriate approach to configuration management. Clear channels and methods of communication nurtured a successful collaborative and cross-functional integration approach. Transitioning swiftly from an externally driven approach to integration, to an inherent, self-orchestrating one across all system areas. This paper presents results as a single point Concept Design for the Integrated Tokamak, a strategic management approach, technological integration challenges (e.g. Services, Maintenance, Neutronics) faced, and considerations for the future (post concept design).Introducing a centralised design process, appropriate control and clear communication utilising CAD has been the key to maturing the STEP Tokamak Concept. Initial focus on facilitating collaboration and utilising this environment as a ‘sandbox’ enabled early establishment of a ‘space-claim’ concept design in which to iterate. Staging the integration approach and CAD model fidelity developments through ‘Identification, Realisation and Accommodation’ built a resilient foundation to progress through the full design process. This approach has enabled early identification of not only where the integration challenges lay but also their significance, resulting in strategic deployment of valuable resource.

Fast ion energy distribution from third harmonic radio frequency heating measured with a single crystal diamond detector at the Joint European Torus.

Neutron spectroscopy measurements with a single crystal diamond detector have been carried out at JET, for the first time in an experiment aimed at accelerating deuterons to MeV energies with radio frequency heating at the third harmonic. Data are interpreted by means of the expected response function of the detector and are used to extract parameters of the highly non-Maxwellian distribution function generated in this scenario. A comparison with observations using a time of flight and liquid scintillator neutron spectrometers is also presented. The results demonstrate the capability of diamond detectors to contribute to fast ion physics studies at JET and are of more general relevance in view of the application of such detectors for spectroscopy measurements in the neutron camera of next step tokamak devices.

Investigation of peeling-ballooning stability prior to transient outbursts accompanying transitions out of H-mode in DIII-D

The H-mode transport barrier allows confinement of roughly twice as much energy as in an L-mode plasma. Termination of H-mode necessarily requires release of this energy, and the timescale of that release is of critical importance for the lifetimes of plasma facing components in next step tokamaks such as ITER. H-L transition sequences in modern tokamaks often begin with a transient outburst which appears to be superficially similar to and has sometimes been referred to as a type-I edge localized mode (ELM). Type-I ELMs have been shown to be consistent with ideal peeling ballooning instability and are characterized by significant (up to ∼50%) reduction of pedestal height on short (∼1 ms) timescales. Knowing whether or not this type of instability is present during H-L back transitions will be important of planning for plasma ramp-down in ITER. This paper presents tests of pre-transition experimental data against ideal peeling-ballooning stability calculations with the ELITE code and supports those results with secondary experiments that together show that the transient associated with the H-L transition is not triggered by the same physics as are type-I ELMs.

Preliminary Thermo-hydraulic and Mechanical Analysis for CFETR Divertor

The China Fusion Engineering Test Reactor (CFETR) is a next step tokamak under designing in China, which should be tritium self-sustained by blanket aimed at getting fusion power of 50–200 MW. One of the most challenging components of CFETR is the actively cooled divertor devoted to controlling the plasma boundary, exhausting the heat and reducing the impurities in the plasma. The thermo-hydraulic and mechanical design of the divertor are particularly demanding because of the high heat loads on the divertor plasma-facing components, the critical heat flux, the pressure drop related to the pumping power and the thermal stresses caused by big temperature gradient. In order to verify the feasibility of the divertor cooling structure, finite element calculations have been carried out. The mass rate of the cooling water should be as high as possible concerning the heat loads bearing of the divertor. But simultaneously the pumping power for the cooling system would be higher. In the paper, an optimal mass rate is given as a compromise between the heat loads bearing and the pumping power based on analysis results.

Diagnosis of physical parameters of fast particles in high power fusion plasmas with high resolution neutron and gamma-ray spectroscopy

High resolution neutron emission spectroscopy (NES) and gamma-ray spectroscopy (GRS) measurements of fast ions in high power fusion plasmas are reviewed. NES is a well established diagnostics of the velocity distribution of fast fuel ions and was recently used to investigate the interaction of energetic ions with MHD instabilities. High energy resolution GRS on fusion plasmas is a more recent application and was shown to provide information on the distribution function of fast minority ions accelerated by ICRH, such as 4He and 3He. Starting from measurements on today's high power D plasmas, fast ion measurements with NES and GRS in a DT burning plasma of next step tokamaks, such as ITER, are discussed. The enhanced neutron and gamma-ray fluxes expected on ITER will allow for time-resolved measurements of the fast fuel and minority ion dynamics in the ms time scale. The intensity of the alpha knock-on component in the 14 MeV neutron spectrum and of the 4.44 and 3.21 MeV gamma-ray peaks from the 9Be(α,nγ)12C reaction is studied as a diagnostics for the α particle slowing down distribution in a DT plasma. The results show that the two techniques are sensitive to different regions of the α particle phase space and thus provide complementary information.

Resonant Pedestal Pressure Reduction Induced by a Thermal Transport Enhancement due to Stochastic Magnetic Boundary Layers in High Temperature Plasmas

Good alignment of the magnetic field line pitch angle with the mode structure of an external resonant magnetic perturbation (RMP) field is shown to induce modulation of the pedestal electron pressure p(e) in high confinement high rotation plasmas at the DIII-D tokamak with a shape similar to ITER, the next step tokamak experiment. This is caused by an edge safety factor q95 resonant enhancement of the thermal transport, while in contrast, the RMP induced particle pump out does not show a significant resonance. The measured p(e) reduction correlates to an increase in the modeled stochastic layer width during pitch angle variations matching results from resistive low rotation plasmas at the TEXTOR tokamak. These findings suggest a field line pitch angle resonant formation of a stochastic magnetic edge layer as an explanation for the q95 resonant character of type-I edge localized mode suppression by RMPs.

Plasma-facing materials for thermo-nuclear fusion devices

The plasma facing materials and components in existing and future fusion devices are strongly affected by plasma wall interaction processes. These components, in particular the first wall (FW), the limiters and the divertor are subject to intense quasi-stationary thermal loads during plasma operation. While the resulting thermal loads to the first wall will remain below 1 MW·m−2, special attention has to be paid to high heat flux components like limiters and the divertor. Here the expected power densities will be at least one order of magnitude above the ones at the FW, with expected peak heat fluxes of up to 20 MW·m−2 for future magnetic confinement devices. Beside quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJ·m−2 are another serious concern for next step tokamak devices, in particular for ITER. The most serious events are plasma disruptions, vertical displacement events, and Edge Localized Modes (so-called ELMs). These requirements make high demands on the selection of qualified materials and reliable fabrication processes for actively cooled plasma facing components. High heat flux test facilities based on intense electron and ion beams have been utilized successfully to assess the efficiency and the fatigue life time of different material solutions and design concepts. Modeling and experiments with both normal operation scenarios and transient events, are being performed to evaluate and to quantify the resulting material erosion or damage and thus to assess the life time of the components. Additional research activities are focused on the degradation of materials and joints due to energetic neutrons. In order to investigate irradiation induced property changes, materials samples and actively cooled plasma facing components have been irradiated in fission reactors and tested in thermal load tests. The technical solutions which are considered today are mainly based on beryllium, carbon materials or tungsten as armor materials and copper alloys or stainless steel for the heat sink. Furthermore, the needs for extensive quality control methods and non-destructive analyses during the procurement phase will be highlighted.

Characterization of the H-mode edge barrier at ASDEX Upgrade

The scaling of the edge transport barrier (ETB) that sustains H-mode plasmas is crucial for the performance of next step tokamaks. At ASDEX Upgrade, the suite of edge diagnostics has been significantly improved to address this issue. High spatial resolution profiles of most of the key edge plasma parameters necessary to determine the magneto hydrodynamic (MHD) stability are now available. New high temporal resolution measurements give clear indications of the nonlinear evolution of the ELM crash. The correlation lengths of edge turbulence have been shown to be correlated with the edge radial electric field shear using a new correlation Doppler reflectometer system. The measured pressure gradient in the ETB is found to be consistent with ideal MHD stability limits, both for Type I and II ELMs. In addition, the edge electron temperature and density gradient lengths are found to be strongly correlated, leaving only the ETB width as a free parameter. In ASDEX Upgrade, the ETB width does not vary significantly over the entire H-mode edge database. Modelling of the transport of comparison discharges in hydrogen and deuterium shows that the expected mass effect on neutral penetration is largely compensated by more efficient heating of deuterium neutrals but requires a transport barrier in both the energy and particle channels in order to reproduce the measured edge temperature and density profiles.

Development of real-time diagnostics and feedback algorithms for JET in view of the next step

Real-time control of many plasma parameters will be an essential aspect in the development of reliable high performance operation of next step tokamaks. The main prerequisites for any feedback scheme are the precise real-time determination of the quantities to be controlled, requiring top quality and highly reliable diagnostics, and the availability of robust control algorithms.A new set of real-time diagnostics was recently implemented on JET to prove the feasibility of determining, with high accuracy and time resolution, the most important plasma quantities. Some of the signals now routinely provided in real time at JET are: (i) the internal inductance and the main confinement quantities obtained by calculating the Shafranov integrals from the pick-up coils with 2 ms time resolution; (ii) the electron temperature profile, from electron cylotron emission every 10 ms; (iii) the ion temperature and plasma toroidal velocity profiles, from charge exchange recombination spectroscopy, provided every 50 ms; and (iv) the safety factor profile, derived from the inversion of the polarimetric line integrals every 2 ms. With regard to feedback algorithms, new model-based controllers were developed to allow a more robust control of several plasma parameters.With these new tools, several real-time schemes were implemented, among which the most significant is the simultaneous control of the safety factor and the plasma pressure profiles using the additional heating systems (LH, NBI, ICRH) as actuators. The control strategy adopted in this case consists of a multi-variable model-based technique, which was implemented as a truncated singular value decomposition of an integral operator. This approach is considered essential for systems like tokamak machines, characterized by a strong mutual dependence of the various parameters and the distributed nature of the quantities, the plasma profiles, to be controlled. First encouraging results were also obtained using non-algorithmic methods like neural networks, which have been successfully applied to non-linear and ill-posed problems, for example the determination of the divertor radiated power.The real-time hardware and software architectures adopted are also described with particular attention to their relevance to ITER.

Neutron measurements on Joint European Torus using an NE213 scintillator with digital pulse shape discrimination

A digital pulse shape discrimination system (DPSD) has been used in conjunction with collimated NE213 scintillators for neutron spectroscopic measurements at high count rates (MHz range) in Joint European Torus discharges (DD and DT fueled, neutral beam injection and rf heated). The system, developed at ENEA-Frascati, is based on a commercial 200 MHz 12-bit analog to digital transient recorder card, which digitizes the direct output signal from the anode of a photomultiplier. Among the unique features of this DPSD system are the possibility of postexperiment data reprocessing, high count rate operation, and simultaneous neutron and gamma (γ) spectroscopy. Separation between γ and neutron (n) events is performed by means of dedicated software exploiting the charge comparison method; separate n and γ pulse height distributions are obtained and an example of neutron spectrum unfolding is shown. Implications of the DPSD in future neutron diagnostic systems on large and next step tokamaks are discussed.

The beta scaling of energy confinement in ELMy H-modes in JET

The disagreement between the weak dependence of the energy confinement time on normalized pressure, β, observed in dedicated scans and the strongly negative dependence in the confinement scaling laws used for the design of next step tokamaks and future reactors, remains an outstanding problem. As such, scans of β have been undertaken in single null, low triangularity (δ ≈ 0.2) ELMy H-mode plasmas in JET with the MarkIIGB-SRP divertor. The scans varied β by a factor of 2.8 (normalized β from 0.72 to 2.04) and covered a range of magnetic fields (1.5–2.3 T), plasma currents (1.5–2.75 MA) and safety factors (q95 = 2.8 and 3.3). A weak β dependence was observed both globally (B0τE varied less than 9% across any one scan) and locally. A scan within Type I ELMy H-modes suggests that this weaker dependence is not due to ELM regimes. A statistical analysis indicates that these results are consistent with log–linear regressions performed on a wide JET database of ELMy H-modes, if correlations in this database are considered.

Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the β limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, Er, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge Er well. The Er at the bottom of the well in these cases is about −100 kV m−1 compared with −20 to −30 kV m−1 in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal β and ν* values matching the values desired for ITER have been achieved. In these higher density plasmas, the Er well is significantly shallower and broader.

Real-time calculation of plasma parameters for feedback control in JET

Plasma parameters, like the internal inductance, li, and the diamagnetic poloidal beta, βDIA, are of particular relevance for a reliable real-time control system of next step tokamaks. These and other quantities have been obtained at Joint European Torus (JET) with a method that uses the Shafranov integrals S1, S2 and S3. Indeed they allow the direct calculation of the Shafranov parameter Λ = βMHD + li/2 and of βDIA. Moreover, in discharges with a sufficiently high elongation (k > 1.3 typically), the internal inductance can be separated from the MHD poloidal beta, βMHD, and calculated independently, through the Shafranov integrals, with a precision that is more than satisfactory for real-time applications. It is worth mentioning that, since S1, S2 and S3 are integrals defined on the plasma boundary, a specific algorithm, depending on the fast code XLOC, has been expressively developed to determine this quantity.The method to determine the plasma parameters has been verified off-line, benchmarking its outputs with the estimates of the most sophisticated equilibrium codes available. The results of this systematic comparison have been very encouraging both in the limiter and x-point phases of the discharges and on all the investigated plasma configurations. The computational time, necessary to determine the plasma boundary and about 50 signals is only about 1.5 ms on a PC equipped with a 400 MHz Pentium II, well below the 10 ms constraint of JET real-time applications. The code has therefore been implemented on-line, and its outputs have already been exploited to achieve the feedback control of some plasma parameters.

Real-time determination of confinement parameters in JET

The main confinement parameters, like the internal inductance l i and the diamagnetic poloidal beta β DIA , are of particular relevance for a reliable real-time control system of next step tokamaks. These quantities have been obtained at Joint European Torus (JET), with a precision more than satisfactory for real-time applications, through a method, known as BETALI, that uses the Shafranov integrals S 1, S 2 and S 3. Since S 1, S 2 and S 3 are defined on the plasma boundary, a technique, that exploits the real-time boundary code XLOC, has been expressively developed to determine the last closed flux surface (LCFS). BETALI has been verified on several experimental plasma configurations, giving very encouraging results both in the limiter and x-point phase of the discharges. The compatibility with the time restrictions has also been tested successfully. This application has, therefore, been implemented and it has already been used during last JET campaigns.

Numerical studies of edge localized instabilities in tokamaks

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code.

Comparison of optimized ECCD for different launch locations in a next step tokamak reactor plasma

Utilization of electron cyclotron radiofrequency sources for current drive and thestabilization of neoclassical tearing modes (NTMs) in next step devices rests on the current density and current drive efficiency attainable. Optimization is reported of electron cyclotron driven current density and current drive efficiency with respect to source frequency as well as toroidal and poloidal launch angles for two launcher positions: an `upper port' position above the midplane and a `midplane' position at the heightof the magnetic axis. The plasma parameters were chosen to be representative of the next step towardsa fusion reactor (B = 5.3 T, ⟨ ne⟩ = 1.0 × 1020m-3, R = 6.2 m). Themodelling is performed with the TORAY code, launching a cone of electron cyclotron rays to account for a finite angular spectrum of injected waves. Current drive at each location is obtained with the Cohen linear model, which includes the effects of toroidal trapping, relativity and wave polarization. For the two launch locations, optimized central current drive efficiency is approximately equal. At plasma radius ρ = 0.835a for stabilization of the q = 2 NTM, the current density is 2.3 times greater and the integrated efficiency is 1.5 times greater for upper port launch relative to midplane launch. A broader range of frequencies for good current drive efficiency is obtained for upper port launch, and this will be reflected by a broader range of operating magnetic field at fixed electron cyclotron source frequency. Twenty megawatts of electron cyclotron power satisfy the criteria forcontrol of NTMs.

Tokamak dust, characterisation and removal studies

Dust and flakes are generated during operation of next step tokamaks such as ITER by plasma interaction with plasma facing components. Accumulation of this particulate engenders potential safety problems, the main ones being chemical reactivity (hydrogen), mobilisation of radioactive dust in a loss of vacuum event, and tritium retention within the plasma chamber. The present prediction for the annualised particulate generation rate is many times more than the currently foreseen safety limits for in-vessel dust inventory. Thus, means may have to be provided to remove particulate on a routine basis. If a vacuum break is needed for batch particulate removal, the resulting reduction in tokamak availability may be unacceptable. Ideally, means should be provided to collect and remove particulate without breaking the vacuum, on a quasi-continuous, intra-shot basis. Vibratory conveying (VC) is one approach to particulate removal. This paper reports initial characterisation of surrogate materials used to emulate tokamak particulate, preliminary results from a small-scale VC test rig, and concepts for implementing VC in a tokamak.

Plasma vertical control with internal and external coils in next step tokamaks

Vertical stability and control are examined for a tokamak configuration intended to be a generic representation of next step devices. Vertical stability calculations show that a critical resistive wall location can be determined for realistic structures and that the introduction of small amounts of low resistivity material to an all steel structure can significantly reduce the vertical instability growth rate. Vertical control simulations show that internal control coils require significantly less feedback power than external coils, and that low resistivity materials can allow either very low feedback powers or the coils to be located externally with reasonable feedback powers.

Recent results from divertor and scrape-off layer studies at JET

Recent progress in the study of divertor and scrape-off layer plasma phenomena inJET is reviewed. Three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have beeninstalled and exploited under reactor relevant conditions. With increased divertor closure,the particle exhaust rate has increased and neutral compression factors of >100 have beenobtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 have been measuredunder a wide range of conditions and satisfy the minimum requirements for ITER. Excellentpower handling has been demonstrated in all divertor configurations. Fast infrared camerameasurements exhibited broad deposition profiles during type I ELMs and energy densities of∼0.12 MJ/m2, which extrapolate to unacceptable levels for ITER. During the recentDT experiments, the co-deposition of tritium on cold shadowed surfaces in the inner divertorwas identified as an important form of long term tritium retention. This has serious implicationsfor the divertor design and tritium inventory in a next step tokamak. Core plasma purity hasnot improved with enhanced divertor closure or decreased main chamber neutral pressure.Studies of the chemical sputtering yield showed a dependence on surface temperature andhydrogen isotope. This accounts for the observation of increased carbon impurity productionin Mark II (at 500 K) relative to Mark I (at 300 K). The variation of the disruptive (L mode)density limit appears to be more strongly dependent on the local impurity production ratethan on the divertor geometry. Significant progress has been made in the study of divertordetachment, and volume recombination has been spectroscopically identified. With increasingisotope mass, detachment and the disruptive density limit occur at lower main plasma density,as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGBdivertor, it was possible to modify the in-out asymmetry of the divertor plasma parametersand energy deposition for the first time.

Progress with heating and current drive technologies

High power heating and current drive (H&CD) facilities are required to fulfil a number of functions in Next Step tokamaks such as ITER and in power plants. The main functions of the ITER H&CD systems are: start-up assist; H-mode access; heating to ignition; supplementary heating for driven-burn and for controlled plasma shut-down; current drive for extended burn and access to steady-state regimes; performance optimisation, control of plasma instabilities and making the plasma rotate. This paper reviews recent progress with H&CD systems, which has largely come about in response to the needs of ITER [1–3]. We survey how the designs have been influenced by physics requirements and technical constraints (such as reliable c.w. operation with modular integrated port plug designs), and emphasise key outstanding issues. The four main candidate systems are: electron and ion cyclotron resonance frequency, lower hybrid, and neutral beam heating. These all have their own physics and technology advantages, and disadvantages, and their components are at various stages of development.