Search

A physics and engineering analysis of alternative divertor configurations is carried out by examining benefits and problems by comparing the baseline single null solution with a Snowflake, an X- and a Super-X divertor. It is observed that alternative configurations can provide margin and resilience against large power fluctuations, but their engineering has intrinsic difficulties, especially in the balance between structural solidity and accessibility of the components and when the specific poloidal field coil positioning poses further constraints. A hybrid between the X- and Super-X divertor is proposed as a possible solution to the integration challenge.

The divertor configuration defines the power exhaust capabilities of DEMO as one of the major key design parameters and sets a number of requirements on the tokamak layout, including port sizes, poloidal field coil positions, and size of toroidal field coils. It also requires a corresponding configuration of plasma-facing components (PFCs) and a remote handling scheme to be able to handle the cassettes and associated in-vessel components (IVC) the configuration requires., There is a risk that the baseline ITER-like single-null (SN) divertor configuration cannot meet the PFC technology limits regarding power exhaust and first wall protection while achieving the target plasma performance requirements of DEMO or a future fusion power plant. Alternative magnetic configurations - double-null, snowflake, X-, and super-X - exist and potentially offer solutions to these risks and a route to achievable power handling in DEMO. But these options impose significant changes on machine architecture, increase the machine complexity and affect remote handling and plasma physics and so an integrated approach must be taken to assessing the feasibility of these options., In this paper we describe the work programme to assess the requirements for incorporating these configurations into DEMO.

A physics and engineering analysis of alternative divertor configurations is carried out by examining benefits and problems by comparing the baseline single null solution with a Snowflake, an X- and a Super-X divertor. It is observed that alternative configurations can provide margin and resilience against large power fluctuations, but their engineering has intrinsic difficulties, especially in the balance between structural solidity and accessibility of the components and when the specific poloidal field coil positioning poses further constraints. A hybrid between the X- and Super-X divertor is proposed as a possible solution to the integration challenge.

Plasma exhaust has been identified as a major challenge towards the realisation of magnetic confinement fusion. To mitigate the risk that the single null divertor (SND) with a high radiation fraction in the scrape-of-layer (SOL) adopted for ITER will not extrapolate to a DEMO reactor, the EUROfusion consortium is assessing potential benefits and engineering challenges of alternative divertor configurations. Alternative configurations that could be readily adopted in a DEMO design include the X divertor (XD), the Super-X divertor (SXD), the Snowflake divertor (SFD) and the double null divertor (DND). The flux flaring towards the divertor target of the XD is limited by the minimum grazing angle at the target set by gaps and misalignments. The characteristic increase of the target radius in the SXD is a trade-off with the increased TF coil volume, but, ultimately, also limited by forces onto coils. Engineering constraints also limit XD and SXD characteristics to the outer divertor leg with a solution for the inner leg requiring up-down symmetric configurations. Capital cost increases with respect to a SND configuration are largest for SXD and SFD, which require both significantly more poloidal field coil conductors and in the case of the SXD also more toroidal field coil conductors. Boundary models with increasing degrees of complexity have been used to predict the beneficial effect of the alternative configurations on exhaust performance. While all alternative configurations should decrease the power that must be radiated in the outer divertor, only the DND and possibly the SFD also ease the radiation requirements in the inner divertor. These decreases of the radiation requirements are however expected to be small making the ability of alternative divertors to increase divertor radiation without excessive core performance degradation their main advantage. Initial 2D fluid modeling of argon seeding in XD and SFD configurations indicate such advantages over the SND, while results for SXD and DND are still pending. Additional improvements, expected from increased turbulence in the low poloidal field region of the SFD also remain to be verified. A more precise comparison with the SND as well as absolute quantitative predictions for all configurations requires more complete physics models that are currently only being developed.

Sedlak, K. et al., The European DEMO, i.e. the demonstration fusion power plant designed in the framework of the Roadmap to Fusion Electricity by the EUROfusion Consortium, is approaching the end of the pre-conceptual design phase, to be accomplished with a Gate Review in 2020, in which all DEMO subsystems will be reviewed by panels of independent experts. The latest 2018 DEMO baseline has major and minor radius of 9.1 m and 2.9 m, plasma current 17.9 MA, toroidal field on the plasma axis 5.2 T, and the peak field in the toroidal-field (TF) conductor 12.0 T. The 900 ton heavy TF coil is prepared in four low-temperature-superconductor (LTS) variants, some of them differing slightly, other significantly, from the ITER TF coil design. Two variants of the CS coils are investigated—a purely LTS one resembling the ITER CS, and a hybrid coil, in which the innermost layers made of HTS allow the designers either to increase the magnetic flux, and thus the duration of the fusion pulse, or to reduce the outer radius of the CS coil. An issue presently investigated by mechanical analyzes is the fatigue load. Two variants of the poloidal field coils are being investigated. The magnet and conductor design studies are accompanied by the experimental tests on both LTS and HTS prototype samples, covering a broad range of DC and AC tests. Testing of quench behavior of the 15 kA HTS cables, with size and layout relevant for the fusion magnets and cooled by forced flow helium, is in preparation., This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. This work was supported in part by the Swiss National Science Foundation (SNF) under contract number 200021_179134.

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 R and in terms of toroidal magnetic field BT. Finally, an attempt to extend the EU-DEMO related conclusions to a more general level is made.

A large scale program to develop a conceptual design for a demonstration fusion power plant (DEMO) has been initiated in Europe. Central elements are the baseline design points, which are developed by system codes. The assessment of the credibility of these design points is often hampered by missing information. The main physics and technology content of the central European system codes have been published (Kovari et al 2014 Fusion Eng. Des. 89 3054–69, 2016 Fusion Eng. Des. 104 9–20, Reux et al 2015 Nucl. Fusion 55 073011). In addition, this publication discusses key input parameters for the pulsed and conservative design option EU DEMO1 2015 and provides justifications for the parameter choices. In this context several DEMO physics gaps are identified, which need to be addressed in the future to reduce the uncertainty in predicting the performance of the device. Also the sensitivities of net electric power and pulse duration to variations of the input parameters are investigated. The most extreme sensitivity is found for the elongation ( Δ κ 95 = 10 % corresponds to Δ P e l , n e t = 125 % ).

This paper describes the status of the pre-conceptual design activities in Europe to advance the technical basis of the design of a DEMOnstration Fusion Power Plant (DEMO) to come in operation around the middle of this century with the main aims of demonstrating the production of few hundred MWs of net electricity, the feasibility of operation with a closed-tritium fuel cycle, and maintenance systems capable of achieving adequate plant availability. This is expected to benefit as much as possible from the ITER experience, in terms of design, licensing, and construction. Emphasis is on an integrated design approach, based on system engineering, which provides a clear path for urgent R&D and addresses the main design integration issues by taking account critical systems interdependencies and inherent uncertainties of important design assumptions (physics and technology). A design readiness evaluation, together with a technology maturation and down selection strategy are planned through structured and transparent Gate Reviews. By embedding industry experience in the design from the beginning it will ensure that early attention is given to technology readiness and industrial feasibility, costs, maintenance, power conversion, nuclear safety and licensing aspects.