Rudolf Neu

The Divertor Tokamak Test (DTT) is a compact sized (R = 2.19 m, a = 0.70 m) fully superconducting tokamak presently under construction at the ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) research center in Frascati, Italy. With its high field (B = 5.85 T) and plasma current (Ip = 5.5 MA), actively cooled in-vessel W components to sustain long pulse duration (tpulse = 100 s), and a significant amount of external heating power (up to P = 45 MW to the plasma), it aims at exploring heat exhaust solutions in fusion-relevant environments by reproducing the level of divertor heat loads expected in ITER and DEMO. Recently, a first version of the DTT research plan was released, defining the project objectives and describing targeted plasma scenarios. The research plan also sets the basis for the development of diagnostic systems, closely following the experimental phases of the machine therein described, with a particular focus toward the divertor region. Since DTT’s first plasma will soon be followed by an experimental phase, operations will start with a full set of systems for machine protection and plasma control, as well as diagnostics to probe initial scenarios. Advanced systems for the investigation of the physics of transport and fast particles will be developed and installed as the machine performance evolves, following its power upgrades. This paper gives an overview of the DTT tokamak targeted diagnostic set, focusing on how these systems are designed to address the scientific goals of the project from first plasma to full power scenarios, and providing context for further publications dedicated to specific diagnostic systems.

The SOLPS-ITER edge plasma numerical model is employed to predict the exhaust operational space in a small-scale tokamak device currently being investigated as the proposed European Volumetric Neutron Source (EU-VNS) with large power-to-size ratio of P / R ∼ 20 MW m − 1 . Previous work is extended by allowing krypton seeding to dissipate and redistribute heat instead of argon. The found finite operational space is limited by both, engineering limits for the peak heat-flux expected at the divertor plasma facing components < 10 MW m − 2 , and at the same time requiring a low enough effective charge Z eff and high electron temperature T e in the core region to keep up the neutron wall load NWL ∼ T e 3 / 2 / Z eff 2 from beam-target fusion. With krypton seeding the operational space seems to be larger than with argon and at the same time the amount of tritium throughput can be halved compared to the previous study. The required total T-throughput rate is of the order 3 − 4 ⋅ 10 22 s − 1 and it is observed that pellet fuelling (neglecting any dynamics involved during the ablation phases) is deemed to be as efficient and required to fuel both the core density and the separatrix density at the same time. A positive correlation is found between the krypton enrichment in the divertor and the increase of upstream density effectively fuelled by pellets. A low level of Z eff ≈ 1.6 , and the observation of the required Greenwald fraction of 50% to achieve this, is compatible with the predictions of the core plasma conditions for the current EU-VNS design point from ASTRA/METIS. Within the model assumptions made in SOLPS-ITER, the observation of a depression of the core electron temperature below 1 keV is critical to maintain the NWL as anticipated from the ASTRA/METIS simulations and thus potentially requires more ECRH power in EU-VNS driving the overall operating costs.

Efficient removal of extreme heat loads from plasma-facing components (PFCs) remains one of the key thermal-hydraulic challenges in fusion technology. In water-cooled divertors, operation in the convective boiling regime offers high heat removal capability, but predictive modelling remains limited by insufficient understanding of boiling under one-sided heating and high coolant velocities, resulting in conservative margins to the critical heat flux (CHF). The present paper describes a physics-guided strategy for developing predictive two-phase computational fluid dynamics (CFD) models for such conditions. The strategy combines three tightly connected elements: boiling mechanism identification through a scaled visualisation experiment, mechanistic closure development within a two-fluid CFD framework, and staged validation against realistic high heat flux (HHF) experiments. The experimental basis includes the FEDORA (Fusion Experiment for Divertor Optimization Research Applications) visualisation experiment, designed to provide local boiling observables in the flow channel under scaled divertor-relevant conditions, and the HHF facilities, which provide realistic validation environments. The current status of model development is most advanced at the level of wall heat flux partitioning. Our recent analyses show that classical quenching-based wall boiling formulations become physically inconsistent under fusion-relevant high flow velocity conditions, whereas a formulation based on sliding-conduction provides a more credible physical basis. This study summarizes the first assessment of that wall boiling framework across a broad fusion-relevant database, together with its first CFD implementation and assessment against HHF experiments in the HADES facility. The results support the proposed development direction, showing that improved wall boiling physics can substantially improve thermal predictions under divertor-relevant conditions, while bulk flow modelling and near-CHF vapour-structure transition remain the next required steps toward a fully predictive divertor-relevant boiling model.

The Divertor Tokamak Test facility (DTT) is a device presently under construction at the ENEA site in Frascati (Italy) in the framework of a joint public/private partnership. It has been designed as a superconducting tokamak with breakeven class performance, with the main objective of developing credible solutions for heat and particle exhaust, a key challenge in view of future fusion reactors. This needs to be addressed in a core-edge integrated approach, to assess the compatibility of exhaust solutions with reactor relevant core performance. In this paper, an overview is provided of the DTT research plan, recently developed by an international team. It covers the DTT programmatic objectives, research strategy and expected scientific contributions connected with the device characteristics, not only in the key area of heat exhaust and edge plasma physics, but also on other subjects of high fusion relevance, such as MHD stability in high performance scenarios, transport and turbulence, energetic particle physics, validation of advanced theoretical developments, as well as tests of technological solutions for reactor relevant components.

The recent transition of the ITER actively cooled First Wall Panel (FWP) from beryllium to tungsten (W) armor tiles necessitates the requalification of the component under high heat flux (HHF) testing. A soft copper (Cu) interlayer is utilized to accommodate the significant thermal expansion mismatch between the W tiles and the underlying CuCrZr heat sink. This study investigates the low cycle fatigue (LCF) performance of the Cu compliance layer using a thermomechanical elastoplastic finite element simulation in Ansys APDL. The simulation models a single tile unit cell subjected to 2 MW/m² HHF cycling, utilizing Chaboche non-linear kinematic hardening to accurately capture strain hardening within the Cu and CuCrZr layers.Results indicate that complete surface-level debonding at the W/Cu interface is likely to occur before the required 30,000 HHF cycles. The LCF life of the Cu interlayer is predominantly limited by the 12 mm W tile thickness and the minimum water coolant temperature of 70°C. Furthermore, a novel cycle jump simulation demonstrates that fatigue fronts develop near debonded corners after approximately 8,000 cycles, significantly accelerating subsequent interfacial failure. Conversely, substituting the Cu interlayer with a W-to-CuCrZr functionally graded material (FGM) eliminated cyclic plastic yielding, presenting a highly effective design alternative for future high heat flux components. Despite not substituting a design-by-experiment approach, these findings provide a quantitative basis for understanding thermomechanical degradation in the ITER flat tile geometry and highlight critical parameters for upcoming experimental validation.