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The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group follows established H-mode scalings (ranging between 2 to 8 mm) while the narrow group falls up to 10 times below these scalings (between 0.2 and 0.8 mm), being comparable to the ion total Larmor radius rather than the ion poloidal Larmor radius. The heat flux profiles of the latter group can only be described by a multi-exponential function, rather than the single exponential function convoluted with a Gaussian. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.

ST40 is a compact, high-field ( $B_{\mathrm{T0}}\unicode{x2A7D} 2.1\,\,\,\textrm{T}$ ) spherical tokamak (ST) with a mission to expand the physics and technology basis for the ST route to commercial fusion. The ST40 research programme covers confinement and stability; solenoid-free start-up; high-performance operating scenarios; and plasma exhaust. In 2022, ST40 obtained central deuterium ion temperatures of $9.6 \pm 0.4\ \textrm{keV}$ , demonstrating for the first time that pilot plant relevant ion temperatures can be reached in a compact, high-field ST. Analysis of these high-ion temperature plasmas is presented, including a summary of confinement, transport and microstability characteristics, and energetic particle instabilities. Recent scenario development activities have focused on establishing diverted H-mode plasmas across a range of toroidal fields and plasma currents, along with scenarios with high non-inductive current fractions. In future operations, beginning in 2025, a 1 MW dual frequency (104/137 GHz) electron cyclotron (EC) system will be installed to enable the study of EC and electron Bernstein wave plasma start-up and current drive. Predictive modelling of the potential performance of these systems is presented.

Ion temperatures of over 100 million degrees Kelvin (8.6 keV) have been produced in the ST40 compact high-field spherical tokamak (ST). Ion temperatures in excess of 5 keV have not previously been reached in any ST and have only been obtained in much larger devices with substantially more plasma heating power. The corresponding fusion triple product is calculated to be ${n_{i0}}{T_{i0}}{\tau _E} \approx 6 \pm 2 \times {10^{18}}{{\text{m}}^{ - 3}}{\text{keVs}}$ . These results demonstrate for the first time that ion temperatures relevant for commercial magnetic confinement fusion can be obtained in a compact high-field ST and bode well for fusion power plants based on the high-field ST.

In reply to the Comment by Biel et al (2016 Nucl. Fusion57 038001) on our recent papers Costley et al (2015 Nucl Fusion55 033001) and Costley (2016 Nucl. Fusion56 066003), we point out that the fusion triple product, nTτE, and fusion power gain, Qfus, cannot be expressed solely in terms of independent engineering design variables such as major radius, R, and toroidal field, B; output performance variables such as normalised beta, βN, safety factor, q, and fusion power Pfus, have to be invoked. Further, we show that the density limit has the effect of largely cancelling the size dependence in nTτE and Qfus, which would otherwise be present, when these parameters are expressed in terms of Pfus. Considerations of engineering aspects are also briefly discussed. [ABSTRACT FROM AUTHOR]

Numerical studies have been made to improve the performance of the central column of a superconducting spherical tokamak fusion pilot plant. The assumed neutron shield includes concentric layers of tungsten carbide and water. The relative thickness of the water layers was varied and a minimum power deposition was found at about 17% of water. It was found advantageous to have an approximately 1.7 times thicker water layer next to the core and a similarly thinner layer next to the plasma. The use of tungsten boride instead of tungsten carbide was shown to make an improvement especially if placed close to the central superconducting core, the inner layer alone reducing the power deposition by 29%. Engineering features such as a central steel tie-bar, an insulating thermal vacuum gap, a wall gap next to the plasma and knowledge of the vertical energy distribution are essential to a successful design and their effects on the power deposition are shown in an appendix. The results have been fitted to model distributions and incorporated into the Tokamak Energy System Code, which can then give predictions of the power deposition as a function of other parameters such as the plasma major radius and the maximum magnetic field permitted on the superconductors. [ABSTRACT FROM AUTHOR]

It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (∼180 MW), high Qfus (∼5), relatively small (∼1.35 m major radius) device are given. [ABSTRACT FROM AUTHOR]

A key challenge in designing a fusion power plant is to manage the heat deposition into the central core containing superconducting toroidal field coils. Spherical tokamaks have limited space for shielding the central core from fast neutrons produced by fusion and the resulting gamma rays. This paper reports a series of three-dimensional computations using the Monte Carlo N-particle code to calculate the heat deposition into the superconducting core. For a given fusion power, this is considered as a function of plasma major radius R0, core radius rsc and shield thickness d. Computations over the ranges 0.6 m ⩽ R0 ⩽ 1.6 m, 0.15 m ⩽ rsc ⩽ 0.25 m and 0.15 m ⩽ d ⩽ 0.4 m are presented. The deposited power shows an exponential dependence on all three variables to within around 2%. The additional effects of source profile, the outer shield and shield material are all considered. The results can be interpolated to 2% accuracy and have been successfully incorporated into a system code. A possible pilot plant with 174 MW of fusion is shown to lead to a heat deposition into the superconducting core of order 30 kW. An estimate of 1.7 MW is made for the cryogenic plant power necessary for heat removal, and of 88 s running time for an adiabatic experiment where the heat deposition is absorbed by a 10 K temperature rise. [ABSTRACT FROM AUTHOR]