Erratum: “Radiative heat load distribution on the EU-DEMO first wall due to mitigated disruptions” (Nuclear Materials and Energy (2020) 25, (S2352179120300971), (10.1016/j.nme.2020.100824))

The publisher regrets for the incorrect affiliation reported in the paper for one of the authors (S. Dulla, Politecnico di Torino). The publisher would like to apologise for any inconvenience caused

Radiative heat load distribution on the EU-DEMO first wall due to mitigated disruptions

The EU-DEMO First Wall (FW) will be a relatively thin structure. In order not to damage this layer, heat loads distributed onto the wall should be carefully controlled. In the case of transient events, as for example plasma disruptions, the steady-state heat load limit (~1-2 MW/m^2) can be largely exceeded for a timespan sufficiently long to cause damages. Therefore, when the control system detects an upcoming disruption, Shattered Pellet Injection (SPI) or Massive Gas Injection (MGI) mitigation techniques can be employed to inject impurities and switch off the plasma safely. In the present work, the Monte-Carlo ray-tracing code CHERAB is used to compute the radiative heat load distribution on the EU-DEMO Plasma Facing Components (PFCs) due to a mitigated plasma disruption. By applying ad-hoc techniques to improve the quality of the Monte Carlo calculation, we obtain a peak radiative load of ~490 MW/m^2 on the PFCs, which is ~25% lower than previous estimates

Divertor currents optimization procedure for JET-ILW high flux expansion experiments

This paper deals with a divertor coil currents optimized procedure to design High Flux Expansion (HFE) configurations in the JET tokamak aimed to study the effects of flux expansion variation on the radiation fraction and radiated power re-distribution. A number of benefits of HFE configuration have been experimentally demonstrated on TCV, EAST, NSTX and DIII-D tokamaks and are under investigation for next generation devices, as DEMO and DTT. The procedure proposed here exploits the linearized relation between the plasma-wall gaps and the Poloidal Field (PF) coil currents. Once the linearized model is provided by means of CREATE-NL code, the divertor coils currents are calculated using a constrained quadratic programming optimization procedure, in order to achieve HFE configuration. Flux expanded configurations have been experimentally realized both in ohmic and heated plasma with and without nitrogen seeding. Preliminary results on the effects of the flux expansion variation on total power radiation increase will be also briefly discussed.EURATOM 63305

The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

In the past at JET, with the MkI divertor, a systematic study of the influence of X-point height and poloidal flux expansion has been conducted [1,2] showing minor differences in the radiation distribution, whereas in [3] experiment and simulations have shown enhancement of detachment as the flux expansion was increased. More recently at JET, equipped with the ITER-like wall (ILW), radiative seeded scenarios have been studied and a maximum radiation fraction 75% has been achieved. EDGE2D-EIRENE [4–6] simula- tions [7,8] have already shown that the divertor heat fluxes can be reduced with N2 injection, qualita- tively consistent with experimental observations [9] , by adjusting the impurity injection rate to reproduce the measured divertor radiation. In this paper we will present edge predictive simulations on the assess- ment of effects of poloidal flux expansion and recycling on radiation distribution and X-point peaking on JET-ILW nitrogen seeded plasmas

Comparison between finite element and experimental evidences of innovative W lattice materials for sacrificial limiter applications

Power exhaust is a key mission for the realization of fusion electricity. Engineering challenges may arise from the extreme heat fluxes developed during plasma transients, above the limit offered by existing materials. These can reduce the lifetime of plasma-facing components (PFCs), imposing extraordinary maintenance, reactor safety issues and ultimately delayed return to normal operation. Concerning the EU DEMO reactor, discrete sacrificial limiters are being investigated as the last safety resource of the reactor's wall in case of unmitigated events. Within this context, micro-engineered tungsten (W) lattices are proposed to cope with unmitigated plasma disruptions. Unlike bulk W, lattices can be tailored to meet the operational requirements of the limiter, compromise between steady-state and off-design performances while avoiding overloading of the heat sink and delay the need for extraordinary maintenance. By calibrating an equivalent solid model originally developed and validated for open-cell aluminum (Al) foams, tailored lattices have been modelled and samples fabricated through additive manufacturing for characterization and testing, currently ongoing. In the present work, the thermal response of lattice samples during thermal shock high heat flux (HHF) tests performed at the linear facility QSPA Kh-50 facility is simulated using ANSYS and compared with available results. Enthalpy changes of W were imposed to simulate phase change. Good agreement with experiments and SDC-IC reference up to melting point was observed. Ultimately, a thermal quench of an unmitigated DEMO disruption was simulated involving an original MAPDL routine that removes mesh elements at the melting or vaporization point.s

Impact of plasma-wall interaction and exhaust on the EU-DEMO design

In the present work, the role of plasma facing components protection in driving the EU-DEMO design will be reviewed, focusing on steady-state and, especially, on transients. This work encompasses both the first wall (FW) as well as the divertor. In fact, while the ITER divertor heat removal technology has been adopted, the ITER FW concept has been shown in the past years to be inadequate for EU-DEMO. This is due to the higher foreseen irradiation damage level, which requires structural materials (like Eurofer) able to withstand more than 5 dpa of neutron damage. This solution, however, limits the tolerable steady-state heat flux to ~1 MW/m2, i.e. a factor 3–4 below the ITER specifications. For this reason, poloidally and toroidally discontinuous protection limiters are implemented in EU-DEMO. Their role consists in reducing the heat load on the FW due to charged particles, during steady state and, more importantly, during planned and off-normal plasma transients. Concerning the divertor configuration, EU-DEMO currently assumes an ITER-like, lower single null (LSN) divertor, with seeded impurities for the dissipation of the power. However, this concept has been shown by numerous simulations in the past years to be marginal during steady-state (where a detached divertor is necessary to maintain the heat flux below the technological limit and to avoid excessive erosion) and unable to withstand some relevant transients, such as large ELMs and accidental loss of detachment. Various concepts, deviating from the ITER design, are currently under investigation to mitigate such risks, for example in-vessel coils for strike point sweeping in case of reattachment, as well as alternative divertor configurations. Finally, a broader discussion on the impact of divertor protection on the overall machine design is presented

Containment structures and port configurations

This article describes the DEMO cryostat, the vacuum vessel, and the tokamak building as well as the system configurations to integrate the main in-vessel components and auxiliary systems developed during the Pre-Conceptual Design Phase. The vacuum vessel is the primary component for radiation shielding and containment of tritium and other radioactive material. Various systems required to operate the plasma are integrated in its ports. The vessel together with the external magnetic coils is located inside the even larger cryostat that has the primary function to provide a vacuum to enable the operation of the superconducting coils in cryogenic condition. The cryostat is surrounded by a thick concrete structure: the bioshield. It protects the external areas from neutron and gamma radiation emitted from the tokamak. The tokamak building layout is aligned with the VV ports implementing floors and separate rooms, so-called port cells, that can be sealed to provide a secondary confinement when a port is opened during in-vessel maintenance. The ports of the torus-shaped VV have to allow for the replacement of in-vessel components but also incorporate plasma limiters and auxiliary heating and diagnostic systems. The divertor is replaced through horizontal ports at the lower level, the breeding blanket (BB) through upper vertical ports. The pipe work of these in-vessel components is also routed through these ports. To facilitate the vertical replacement of the BB, it is divided into large vertical segments. Their mechanical support during operation relies on vertically clamping them inside the vacuum vessel by a combination of obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the breeding blanket structural material in the toroidal magnetic field