The Culham Centre for Fusion Energy (CCFE) is the UK's national laboratory for fusion research. It is located at the Culham Science Centre, near Culham, Oxfordshire, and is the site of the Joint European Torus (JET), Mega Ampere Spherical Tokamak (MAST) and the now closed Small Tight Aspect Ratio Tokamak (START).Formerly known as UKAEA Culham, the laboratory was renamed in October 2009 as part of organisational changes at its parent body, the United Kingdom Atomic Energy Authority (UKAEA).Since 2016, the director has been Professor Ian Chapman, and the centre has been engaged in work towards the final detailed design of ITER as well as preparatory work in support of DEMO.In 2014 it was announced the centre would house the new RACE (Remote Applications in Challenging Environments).
Radiation damage profoundly alters the physicochemical properties of materials subjected to extreme environments. A key question within this context is whether radiation damage accumulation reaches saturation, particularly at elevated temperatures relevant to advanced nuclear energy systems. Employing well-designed irradiation protocols and a suite of advanced characterization techniques, we have experimentally demonstrated damage saturation in the stage III recovery regime in heavy-ion irradiated tungsten-a candidate plasmafacing material expected to operate reliably at/above this baseline temperature in fusion reactors. We validated that damage saturation reflects an asymptotic steady-state, established via the dynamic equilibrium between defect generation and annihilation. This phenomenon represents a limit to damage accommodation capacity in tungsten, independent of irradiation sources, although the onset of damage saturation exhibits a temperaturedependent shift. While observed in BCC tungsten, this phenomenon of damage saturation is not exclusive to this system and is expected to apply broadly across FCC and HCP materials. Therefore, defining these saturation conditions will offer a critical benchmark for assessing the mid- to long-term performance of materials and guiding the design of more radiation-tolerant systems.
This work advances the quantitative validation of SOLPS-ITER against Tokamak & agrave; Configuration Variable (TCV) ohmic L-mode discharges. It shows that the treatment of the ion parallel heat flux limiter and the assumptions adopted for chemical sputtering result in a consistent explanation for the divertor over-cooling and excessive divertor density predicted by earlier simulations. A systematic heat flux limiter scan highlights that a strong restriction of ion heat conduction significantly reduces both electron Te,t and ion target temperatures Ti,t and drives a compensating increase in plasma density ne,t, resulting in an overly cold and dense divertor solution. The availability of new Te and Ti measurements along the outer divertor leg of TCV additionally constrains the model and supports the use of a weakly limited ion heat flux to best reproduce experimental observations. Restricting chemical sputtering to only tiles in the immediate vicinity of the strike points suppressed a spurious low-field-side carbon source associated with thermal molecules and significantly improves agreement with two-dimensional C iii emissivity reconstructions. The resulting SOLPS-ITER setup achieves improved agreement across multiple diagnostics simultaneously for upstream, divertor-leg, and target electron density and temperature profiles, as well as for target heat fluxes and radiation measurements.
This paper gives an overview of erosion and migration studies of tungsten (W) in the WEST tokamak during its Phase 1 (2016–2021) and Phase 2 (from 2022) experimental campaigns with a focus on plasma-facing components (PFCs) at the divertor. In Phase 1, gross erosion of PFCs is in line with observations from other major fusion devices and attributed to low- Z impurities in the plasma. In addition, a strong asymmetry is observed between the high- (inner) and low-field (outer) side divertor targets, in favour of the inner side. Net erosion at rates of <0.5 nm s ^−1 is measured around the strike points while the remaining areas are dominated by net deposition. The thickest deposited layers (up to 50 μ m) with the most complex structures result from a cumulated plasma exposure of ∼7 h. The overall erosion-deposition pattern is further influenced by the strong magnetic ripple of WEST, which can result in almost an order of magnitude difference between the maxima and minima of the ripple. In Phase 2, increasing plasma fluence leads to the deposits growing to hundreds of micrometres in thickness. At the same time, erosion proceeds at a constant rate and can reach values up to 30 µ m in ∼18 h of plasma time. In the main chamber, erosion is weaker than at the divertor but especially at low densities it can result in notable transport of W into the core. In addition, upon switching on the ICRF antennas, W sputtering on the close-by limiter structures can increase by a factor of more than 10. Modelling is able to catch many of the observed phenomena in Phase 1, with the exception of the inner–outer asymmetry and the formation of the thick deposits. In contrast, the patterns during the high-fluence operations in Phase 2 require more work to be reproduced.
An effective fusion reactor maintenance scheme enables safe operations and short downtimes. This in turn leads to high availability, which is critical to the commercial viability of a power-producing plant. In tokamak-based fusion power plants, the chosen maintenance approach has a significant impact on the spatial design of the tokamak, as well as the surrounding infrastructure, and therefore needs to be considered from the outset. Tokamak Energy has developed a pre-concept design of a fusion power plant, ST-E1. This work describes the major drivers and constraints that have been considered, presents the tokamak architecture and chosen maintenance regime, and discusses how this enables the plant’s two-phased approach to demonstrating commercial operations. It also shows the implications for the design of other systems areas, in particular the machine structural arrangement and bioshield and hot cell layout. The reactor core segmentation and removal scheme replaces entire toroidal segments radially through a large vacuum port, along a single axis only. The result is a change-tolerant machine and plant layout that can accommodate the evolving designs of the tokamak.
Operation with beam heating is prone to central accumulation of high- Z impurities in present-day tokamaks due to a dominant neoclassical pinch. Fusion reactors are expected to have a different core transport regime for high- Z impurities, where turbulent transport completely dominates in both diffusion and convection, causing flat impurity profiles. In this work, we present integrated simulations of plasmas that approach this regime of less dominant neoclassical and stronger turbulent high- Z impurity transport in three tokamaks: ASDEX Upgrade, JET and Alcator C-Mod. The modelling is first validated against a suite of diagnostics measuring both the main plasma and the impurity profiles. The high- Z impurity densities are found to be flat in both experiments and simulations. The impurity transport coefficients as calculated by theory-based turbulent and neoclassical transport models are then analysed, showing comparable or even dominant turbulent components instead of the more typical dominant neoclassical high- Z impurity convection. The implications for a reactor are discussed.
