Search

We study numerically the feasibility of achieving the L-H transition in the current EU-DEMO tokamak baseline using uniquely direct electron heating. The ASTRA code coupled to the TGLF turbulent transport model is used to predict steady-state kinetic plasma profiles for diverse numerical scans. Among them, we have varied the separatrix electron density, the total amount of ECRH power, the microwave beam deposition profile and the plasma impurity content. The solutions are then compared to L-H transition scaling laws to assess whether the found plasma state would enter into H-mode. We find the plasma density and impurity content to be the key variables setting the boundaries in parameter space where the L-H transition is feasible. When impurities can be controlled under a certain threshold, given here for a fully shaped DEMO plasma, the L-H transition is achieved in all the studied conditions.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

High temperature superconductors (HTSs) offer the promise of operating at higher magnetic field and temperature. Recently, the use of high field magnets (by adopting HTS) has been promoted by several groups around the world, including new start-up entries, both to substantially reduce the size of a fusion power reactor system and as a breakthrough innovation that could dramatically accelerate fusion power deployment. This paper describes the results of an assessment to understand the impact of using high field magnets in the design of DEMO in Europe, considering a comprehensive list of physics and engineering limitations together with the interdependencies with other important parameters. Based on the results, it is concluded that increasing the magnetic field does not lead to a reduction in device size with relevant nuclear performance requirements, because (i) large structures are needed to withstand the enormous electromagnetic forces, (ii) thick blanket and n-shield structures are needed to protect the coils from radiation damage effects, and (iii) new divertor solutions with performances well beyond today’s concepts are needed. Stronger structural materials allow for more compact tokamaks, but do not change the conclusion that scalability is not favourable when increasing the magnetic field, beyond a certain point, the machine size cannot be further reduced. More advanced structural support concepts for high-field coils have been explored and concluded that these solutions are either unfeasible or provide only marginal size reduction, by far not sufficient to account for the potential of operating at very high field provided by HTS. Additionally, the cost of high field coils is significant at today’s price levels and shows to scale roughly with the square of the field. Nevertheless, it is believed that even when not operated at high field and starting within conventional insulated coils, HTS can still offer certain benefits. These include the simplification of the magnet cooling scheme thanks to increased temperature margin (indirect conduction cooling). This in turn can greatly simplify coil construction and minimize high-voltage risks at the terminals.

In 2021 JET exploited its unique capabilities to operate with T and D–T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D–T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements—new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors—as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D–T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D–T plasmas to date and expanding our understanding of isotopes and D–T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D–T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D–T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D–T operations is presented in (King et al Nucl. Fusion submitted).

One of our grand challenges towards fusion energy is the achievement of a high-performance plasma core coupled to a boundary solution. The high confinement mode (H-mode) provides such a high-performance fusion core due to the build-up of an edge transport barrier leading to a pedestal. However, it usually features type-I edge localized modes (ELMs) which pose a threat for long-duration plasma operation in future fusion devices as they induce large energy fluences onto the plasma facing components and typically are projected to damage the first wall. For future fusion devices, the integration of a stationary no-ELM regime with a power exhaust solution is indispensable. Several no-ELM and small-ELM regimes have extended their operational space in the past years, with the ultimate goal of providing an alternative core–edge solution to ITER and EU-DEMO. Prominent no-ELM or small-ELM alternatives include the I-mode, QH-mode, EDA H-mode, quasi-continuous exhaust (QCE) and ‘grassy’ ELM regimes, X-point radiator scenarios and negative triangularity L-mode. The state-of-the-art, including access conditions and main signatures, of these alternative regimes is reviewed. Many of these regimes partly match the operational space of ITER and EU-DEMO, however, knowledge gaps remain. Besides compatibility with divertor detachment and a radiative mantle, these include extrapolations to high Q operations, low core collisionality, high Greenwald fractions, impurity transport, amongst others. The knowledge gaps and possible strategies to close these gaps to show their applicability to ITER and EU-DEMO are discussed.

A double-null configuration is being considered for the EU-DEMO, due to its potential benefits for power exhaust arising from the use of two active divertors and magnetically disconnected low- and high-field sides. Using systematic parameter scans in fluid simulations, we have investigated the divertor power exhaust in the EU-DEMO in a connected double-null configuration, and compared the edge plasma properties to those obtained in a single-null configuration under detached conditions anticipated for reactor operation. Neglecting drift effects and kinetic behaviour of the neutrals, no clear benefits of the double-null configuration could yet be identified for the radiation pattern and power mitigation on open field lines. Future work should address the aforementioned physics as well as the effect of the additional X-point on core radiation.

In the present work, the role of plasma facing components protection in driving the EU-DEMO design will be reviewed, focusing on steady-state and, especially, on transients. This work encompasses both the first wall (FW) as well as the divertor. In fact, while the ITER divertor heat removal technology has been adopted, the ITER FW concept has been shown in the past years to be inadequate for EU-DEMO. This is due to the higher foreseen irradiation damage level, which requires structural materials (like Eurofer) able to withstand more than 5 dpa of neutron damage. This solution, however, limits the tolerable steady-state heat flux to ~1 MW/m2, i.e. a factor 3–4 below the ITER specifications. For this reason, poloidally and toroidally discontinuous protection limiters are implemented in EU-DEMO. Their role consists in reducing the heat load on the FW due to charged particles, during steady state and, more importantly, during planned and off-normal plasma transients. Concerning the divertor configuration, EU-DEMO currently assumes an ITER-like, lower single null (LSN) divertor, with seeded impurities for the dissipation of the power. However, this concept has been shown by numerous simulations in the past years to be marginal during steady-state (where a detached divertor is necessary to maintain the heat flux below the technological limit and to avoid excessive erosion) and unable to withstand some relevant transients, such as large ELMs and accidental loss of detachment. Various concepts, deviating from the ITER design, are currently under investigation to mitigate such risks, for example in-vessel coils for strike point sweeping in case of reattachment, as well as alternative divertor configurations. Finally, a broader discussion on the impact of divertor protection on the overall machine design is presented.

The design of future fusion reactors and their operational scenarios requires an accurate prediction of the plasma confinement. We have developed a new model that integrates different elements describing the main physics phenomena which determine plasma confinement. In particular, we are coupling a new pedestal transport model, based on empirical observations, to the ASTRA transport code, which, together with the TGLF turbulent transport model and the NCLASS neoclassical transport model, allows us to describe transport from the magnetic axis to the separatrix. We also coupled a simple scrape-off layer model to ASTRA, which provides the boundary conditions at the separatrix, which are a function of the main engineering parameters. By this way no experimental data of the kinetic profiles is needed, and the only inputs of the model are the magnetic field, the plasma current, the heating power, the fueling rate, the seeding rate, the plasma boundary, and the effective charge. In the modeling work-flow, first a scan in pedestal pressure is performed, by changing the pedestal width. Then the pedestal top pressure is determined using the MISHKA MHD stability code. This modeling framework is tested by simulating ASDEX Upgrade discharges. We show comparisons with experimental fueling, power, and current scans. The changes of the pedestal structure and the gradients in the plasma core, due to the different combinations of fueling, heating power, and plasma current, are well captured by the model. We also show that the predicted pedestal and global stored energies are in good agreement with the experimental measurements, and more accurate with respect to the prediction of the IPB98(y,2) scaling law. The long term goal is to obtain a robust and complete model which can be used to identify important hidden dependencies affecting global plasma confinement, which are difficult to capture by statistical regressions on global parameters. [ABSTRACT FROM AUTHOR]

This paper discusses the criteria to be used in the preliminary design phases of the EU-DEMO reactor to ensure the performance of the divertor without compromising the stability of core plasma or the fusion power generation. This work refers to a lower single null conventional divertor using actively cooled solid metal plasma-facing components and with extrinsic seeding for heat flux dissipation, which is the solution currently being adopted for EU-DEMO. The analysis does not consider the role of edge localised modes, and also neglects major off-normal events like disruptions. It is shown that it is necessary to fulfil two high-level requirements, namely: (i) the concentration of seeded impurities has to be lower than some critical value in order to not compromise the fusion plasma performance or stability and (ii) damage to the divertor plate in the case of accidental plasma reattachment must be avoided for a sufficiently long time in order to ensure safe, controlled termination of the plasma discharge. These requirements are needed because in a device like EU-DEMO, other strategies relying on mass injection are considered more likely to cause a loss of plasma stability at full current, with dramatic consequences for the integrity of plasma facing components. Two figures of merit, corresponding to these criteria, have been identified in the existing literature and discussed. The dependence of such figures of merit on the relevant machine parameters (major radius and toroidal magnetic field) is analysed. Initially, the analysis is carried out using a simple 0D physics approach and subsequently by means of the systems code PROCESS, which allows for consideration of further parameters, such as aspect ratio and elongation. The main conclusion of the present work is that the simultaneous fulfilment of both requirements limits the viable EU-DEMO size both in terms of major radius and in terms of toroidal magnetic field . Finally, an attempt to extend the EU-DEMO related conclusions to a more general level is made. [ABSTRACT FROM AUTHOR]

In the European Union strategy, DEMO is the single step between ITER and a fusion power plant. The current state of physical knowledge and technological limits does not yet well allow us to identify the best combination of solutions to decide the architecture of this new tokamak. To this purpose, an important element of tokamak research finalized to the prediction, the interpretation and the planning of experiments is represented by the ‘integrated modeling’. In this work we use the integrated modeling ASTRA tool coupled with the TGLF transport model in order to investigate scenarios of interest for DEMO. We obtain scaling laws which describe and serve as a guide-line for the pre-conceptual design of the machine. In particular, by means of scaling laws we investigate the performances of the reactor in terms of fusion power, confinement time H-factor and fusion power multiplication factor Q, in different regimes such as H-mode, I-mode and L-mode. In this way we provide important key points for the establishment of the best physical and technological solutions for DEMO scenarios. [ABSTRACT FROM AUTHOR]

The design and optimization of a future tokamak fusion reactor, from the point of view of the plasma performance and output fusion power, requires the understanding of the underlying physics and the validation on present experiments of the theory-based tools used for the predictions. Present experimental efforts are devoted to approach reactor-relevant parameters (collisionality, , normalized heat fluxes and sources, temperatures ratio) as much as possible. In this work a series of discharges performed on ASDEX Upgrade are presented, where plasma density and auxiliary power levels are scanned to cover a parameter space that moves towards reactor relevant parameters. On this dedicated discharge dataset, a first-principle-based model is then applied for validation. The degree of agreement between code results and experimental measurements is shown and discussed. In cases where discrepancy is found, possible causes are identified. It is shown that the employed modeling tool can predict the overall trend, consistent with more fundamental theoretical considerations, although quantitative extrapolation still has to be performed with care. Key results of this work show that the density peaking is mainly sustained by turbulence, with a minor role for the fueling source, and how this experimental demonstration is important to predict future reactor density profiles. Moreover, the role of electromagnetic effects is also pointed out. [ABSTRACT FROM AUTHOR]

During the pre-conceptual design phase of fusion devices such as the European demonstration fusion power plant (DEMO), systems codes provide a fast evaluation of optimal design points and highlight high impact areas. However, determining or evaluating a design point at such an early stage comes with uncertainties in many of the design parameters. These uncertainties are both associated with the physics as well as the engineering basis of the European DEMO design. The work applies an uncertainty quantification analysis to the 2017 pulsed European DEMO design using the PROCESS systems code. It assumes that DEMO will be built as suggested by the baseline and explores what implications the currently known physics and engineering uncertainties have on the expected performance parameters (net electric output and pulse length), while optimising the fusion gain Q. A more detailed single parameter analysis clearly identifies high impact parameters. This confirms previous investigations as well as revealing new areas that warrant deeper investigation in particular in the technology area. [ABSTRACT FROM AUTHOR]

In the present work, the tolerable impurity level and composition for a reactor plasma using several sets of model assumptions are evaluated. Special care was taken to evaluate a comprehensive and consistent set of atomic data for 35 different elements, such that the impurity level for various elements may be studied as a function of their nuclear charge. The data set may not only be useful for the presented work or for system codes which design fusion reactors, but also for interpretation of bolometric measurements. Additionally, the predictions of the spectral distribution of the radiated power is of high quality such that soft x-ray broadband measurements may be interpreted. In the present work the data is used for predicting the radiated power in a reactor plasma, using a 0D, several variants of a 0.5D model and a realistic 1D ASTRA modelling of a DEMO plasma, i.e. the EU DEMO1 2015 design. The maximal or appropriate impurity content of a reactor plasma for all models can be determined, such that the predictions from a simplistic 0D model can be compared to less simplistic models and a proper reactor simulation. These comparisons suggest that with the simplistic models the impurity content may be estimated within a factor of about 1.5, independent of the realization of the reactor plasma. At the same time this study underlines the sensitivity of the reactor performance on the impurity mixture and especially of the He content of the plasma. Additionally, an extended 0.5D model is presented which is able to predict variations of the fusion yield Q and the He concentration, when both is known for a reference scenario. These predictions prove to be of high accuracy when compared to the 1D ASTRA modelling and thus, allow the net impact of an increased dilution and a simlutaneous temperature rise at constant plasma pressure to be evaluated. Furthermore, the parameter space is scanned with more than 105 model reactor plasmas demonstrating that the use of a low-Z impurity diminishes the possibility of an economical feasible reactor plasma. The main results of the parameter scan are made available via scaling formulae. [ABSTRACT FROM AUTHOR]

The upstream separatrix electron density is an important interface parameter for core performance and divertor power exhaust. It has been measured in ASDEX Upgrade H-mode discharges by means of Thomson scattering using a self-consistent estimate of the upstream electron temperature under the assumption of Spitzer–Härm electron conduction. Its dependence on various plasma parameters has been tested for different plasma conditions in H-mode. The leading parameter determining ne,sep was found to be the neutral divertor pressure, which can be considered as an engineering parameter since it is determined mainly by the gas puff rate and the pumping speed. The experimentally found parameter dependence of ne,sep, which is dominated by the divertor neutral pressure, could be approximately reconciled by 2-point modelling. [ABSTRACT FROM AUTHOR]

An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field BT—for different safety factors q, allowable loop voltages Vloop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion53 073019 and Wenninger et al 2017 Nucl. Fusion57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and BT once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L–H transition) is present. [ABSTRACT FROM AUTHOR]

We present an approach to design in a consistent way a stepladder connecting ITER, DEMO and an FPP, starting from an attractive FPP and then locating DEMO such that main similarity parameters for the core scenario are constant. The approach presented suggests how to use ITER such that DEMO can be extrapolated with maximum confidence and a development path for plasma scenarios in ITER follows from our approach, moving from low βN and q typical for the present Q  =  10 scenario to higher values needed for steady state. A numerical example is given, indicative of the feasibility of the approach, and it is backed up by more detailed 1.5-D calculation using the ASTRA code. We note that ideal MHD stability analysis of the DEMO operating point indicates that it is located between the no-wall and the ideal wall β-limit, which may require active stabilization. The DEMO design could also be a pulsed fallback solution should a stationary operation turn out to be impossible. [ABSTRACT FROM AUTHOR]

A summary of recent results on filamentary transport, mostly obtained with the ASDEX-Upgrade tokamak (AUG), is presented and discussed in an attempt to produce a coherent picture of scrape-off layer (SOL) filamentary transport. A clear correlation is found between L-mode density shoulder formation in the outer midplane and a transition between the sheath-limited and the inertial filamentary regimes. Divertor collisionality is found to be the parameter triggering the transition. A clear reduction of the ion temperature takes place in the far SOL after the transition, both for the background and the filaments. This coincides with a strong variation of the ion temperature distribution, which deviates from Gaussianity and becomes dominated by a strong peak below 5 eV. The filament transition mechanism triggered by a critical value of collisionality seems to be generally applicable to inter-ELM H-mode plasmas, although a secondary threshold related to deuterium fueling is observed. EMC3-EIRENE simulations of neutral dynamics show that an ionization front near the main chamber wall is formed after the shoulder formation. Finally, a clear increase of SOL opacity to neutrals is observed, associated with the shoulder formation. A common SOL transport framework is proposed to account for all these results, and their potential implications for future generation devices are discussed. [ABSTRACT FROM AUTHOR]

A 0D analytical stationary model to estimate the degree of detachment on the divertor plates is derived, starting from previous works of Igitkhanov (Igitkhanov et al 1994 21st EPS (Montpellier) ECA v.18B, 1995 22th EPS (Bournemouth) ECA v.19C, Igitkhanov 2014 KIT Sci. Rep. 7661). It accounts for heat convection, conduction, impurity radiation and contains a simplified balance for the neutrals. The upstream particle flux (or, alternatively, the density upstream nup), the heat flux in the scrape-off layer (SOL) qup and the impurity concentrations are required as input, whereas the temperatures at the plate and at the separatrix are not fixed a priori. The routine has been mainly developed for system codes, or more in general for scopy studies in the framework of the preliminary design phase for the reactor DEMO, its simplicity being therefore justified by the purpose of keeping the calculation times as low as possible. A calibration against a more detailed 1D routine (Kallenbach et al 2016 Plasma Phys. Control. Fusion58 045013) has been performed, finding a reasonable agreement. Furthermore, a coupling of the model with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Fable et al 2013 Plasma Phys. Control. Fusion55 124028) to illustrate its possible usage is also presented. [ABSTRACT FROM AUTHOR]

A finite radial temperature gradient has been observed to be maintained within magnetic islands in gyro-kinetic turbulence simulations despite the fast motion along the field, and is related to the trapped particles which do not move along the field around the island. Recent calculations of the interaction of drift-wave turbulence with magnetic islands have shown that turbulence can exist within the separatrix, which in turn allows only the partial flattening of electron temperature profiles. Here we calculate, using a minimal drift-kinetic model, the effect on the bootstrap current in a tokamak. Consequences for the stability of the neoclassical tearing mode are discussed. [ABSTRACT FROM AUTHOR]