Staying positive: producing net power

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)’.

Plasma burn—mind the gap

Meyer

2024

Managing the heat: In-Vessel Components

Cane,

Barth,

Farrington

et al. 2024

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 (D 2 O) 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)’.

Fusing together an outline design for sustained fuelling and tritium self-sufficiency

Lord,

Bennett,

Harrington

et al. 2024

Ensuring tritium fuel self-sufficiency while maintaining continuous and high-specification fuel flow to the tokamak via a low tritium inventory and controllable fuel cycle is a significant challenge to the STEP plant design. Effective and high-quality fuelling and exhaust design is required to sustain and control a stable plasma, whereas fuel sufficiency is required to prevent depletion of available tritium supply. Concerns regarding the lack of tritium availability preventing continuous tritium import are countered by breeding, where highly energetic neutrons from the core fusion reactions interact with lithium atoms suspended in the surrounding breeder blanket to produce tritium. The compact nature of STEP prohibits the integration of inboard breeder blankets posing a significant challenge for the design team looking to ensure more tritium is bred and made available than consumed within the core plasma. This paper outlines how purposeful technology selection and system architecting has converged on the outline of a conceivable and tritium-capable fuel cycle and breeder blanket design. Before introducing the STEP fuel cycle design outline and summarizing the approach undertaken to address the challenges facing plasma fuelling, key aspects of fuel self-sufficiency are discussed. This includes discussing a proposed helium-cooled liquid lithium breeder blanket and possible technology options for tritium extraction from lithium. Lastly, there is a brief process modelling overview, which emphasizes the central contribution of various employed modelling methods. Reflections on the presented fuel cycle design outline conclude that substantial development work is still required to realize a continuous tritium fuel cycle design and overcome the major challenges posed by tritium and lithium handling. Reflections on the presented breeder blanket design proposal conclude that while many substantial risks and blockers remain to achieve fuel self-sufficiency, high breeding ratios are expected to be achievable with a compact spherical tokamak configuration. Nonetheless, it is recognized that further consideration is required to ensure that the selection of liquid lithium as a breeder medium provides the overall simplest route to a self-sufficient and realizable design. This article is part of the theme issue ‘Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)’ .