Recent Publications

WCLL and HCCP test blanket systems: common technologies and integration in ITER

I. Ricapito, M.Y. Ahn, A. Aiello, B. Bellin, S. Cho, S. D’Amico, M. Fabbri, F. Fantini, E. Garcia, L. Giancarli, et al.

Fracture toughness of European reference RAFM steel EUROFER97

E. Gaganidze, D. Terentyev, M. Serrano, P. Lamagnère, M. Walter, T.T.M. Nguyen, A. Zinovev, D. Bermúdez Parra, G. Pintsuk, G. Aiello, et al.

Status of the 3 MW ECRH System at 105 GHz for WEST

Lena Delpech, Stefano Alberti, Konstantinos Avramidis, Aline Ayreault, Julien Barra, Jean-Michel Bernard, Vincent Bruno, Francis Bouquey, Ioannis Chelis, Frederic Clairet, et al.

Simplified conductance model for the subdivertor of Wendelstein 7-X based on high-fidelity simulations and comparisons with experimental data

V. Haak, L. Steiniger, C. Day, J. Igitkhanov, F. Litovoli, D. Naujoks, V. Perseo, G. Schlisio, H. Strobel, C. Tantos, et al.

Neutronics Analysis of Shielding Performance for VNS Inboard Toroidal Field Coil with Various Material Options

Jin Hun Park, Pavel Pereslavtsev, Roman Afanasenko, Dieter Leichtle

Impact of ferritic steel blankets on the magnetic field configuration of the helical fusion reactor

Miguel Lindero-Hernández, Luis Alvaro Montoya-Santiyanes, Nagato Yanagi, Frederic Trillaud, Gabriel dos Santos, Diego A. Garfias-Dávalos, Yoshiro Narushima

Wire electron beam additive manufacturing for monoblock HCPB breeding blanket components: Preliminary assessment through experimental investigation

Gaurav Verma, Luděk Stratil, Jörg Rey, Bernd Baufeld, Martin Petersen, Stephen Fischer, Heiko Neuberger, Francisco A. Hernández, Jarir Aktaa

ITER water cooled lithium lead Test Blanket System transient thermal-hydraulic analysis

Tommaso Glingler, Gianfranco Caruso, Danilo Nicola Dongiovanni, Salvatore d’Amico, Italo Ricapito, Paul Wouters, Francesca Fantini, Joelle Elbez-Uzan, Sandrine Rosanvallon, Matteo D'Onorio

The planned experimental activities of the ITER Test Blanket Module program will be a key step in understanding the physics and technologies required for tritium breeding blankets, which are essential for achieving tritium self-sufficiency in a demonstration fusion power reactor. The ITER Test Blanket Module program foresees the installation and operation of different breeding blanket concepts, such as the Water-Cooled Lithium Lead and the Helium-Cooled Pebble Bed, to demonstrate tritium breeding, high-grade heat extraction, and integrated system operation under ITER conditions. The objective of this paper is to provide an overview of the thermal-hydraulic analyses performed in the frame of EUROfusion, for the Water Cooled Lead Lithium Test Blanket Module, with emphasis on the main phenomena occurring during representative accident sequences. The main figure of merit is that the structural integrity of confinement barriers is preserved. Accident scenarios were analyzed using MELCOR version 1.8.6 for fusion applications. The main sequences considered, were identified through an extensive failure mode and effect analysis, include ex-vessel Loss of Coolant Accident and Loss of Heat Sink. The results highlight the robustness of the current design.

Segregated and block-coupled approaches for electric potential coupling in multi-region magnetohydrodynamic flows within fusion liquid metal breeding blankets

B. Lyu, L. Bühler, C. Mistrangelo

The fixed boundary plasma equilibrium basis for a one Gigawatt electric stellarator power plant

Samuel A. Lazerson, António J. Coelho, Dana Douqa, Adrian Asmund Fessler, Luisa Hübner, Matteo Moscheni, Eoin Hodge, Richard Kembleton, Joona Sissonen, Konsta Särkimäki, et al.

A fixed boundary stellarator equilibrium capable of producing 3 GW of fusion power (1 GW-electric) is presented as the design basis for the GIGA fusion power plant being developed by Gauss Fusion GmbH. The stellarator concept provides a steady-state, transient free, low recirculating power approach to a fusion power plant, which builds on 50 years of progress in plasma physics. A set of requirements for a fixed boundary equilibrium were determined through application of 0.5 D modeling. Optimization of a modified Wendelstein 7-X (W7-X) equilibrium was performed to achieve these requirements including alpha power confinement greater than $85\%$, neoclassical effective ripple below 0.01, bootstrap current below 50 kA, and reduced turbulent heat fluxes. In order to fix the plasma volume of $1500~m^3$ during optimization, the VMEC code was modified to renormalize the boundary coefficient to the desired plasma volume. The STELLOPT stellarator optimization code was modified as well to include new bootstrap current targets, a new target for the radial electric field, and the capability to hold the magnetic field on axis at a fixed value. An intermediary conceptual design plasma and final evolved fixed boundary equilibria are compared to the original modified W7-X equilibrium. The final evolved equilibrium is shown to achieve all the necessary requirements for the GIGA fusion power plant through more detailed modeling of stability, fast ion confinement, and transport.

IFMIF-DONES concrete neutron shielding benchmarks

Haridev Chohan, Martin Ansorge, Yuefeng Qiu, Tomasz Piotrowski, María José Martínez-Echevarría Romero, Ocean Wong, Ivan Kodeli, Joseph Neilson, Kimberley Lennon, Robin Smith, et al.

Performance of the W7-X ECRH gyrotron diamond output window at 2 MW operation

G. Aiello, B. Gorr, J. Jelonnek, H.P. Laqua, F. Legrand, A. Meier, T. Scherer, S. Schreck, D. Strauss, M. Thumm

Thermal behaviour of ZrCo getter beds during hydrogen absorption at variable flow rates and concentrations

Nicolas Bekris, Monica Sirch, Robin Größle