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Gaurav Verma, Luděk Stratil, Jörg Rey, Bernd Baufeld, Martin Petersen, Stephen Fischer, Heiko Neuberger, Francisco A. Hernández, Jarir Aktaa
Xichen Su, Rendeng Tang, Fujun Gou, jianjun Wei, Zongbiao Ye
G.H. Kim, Y.J. Jung, M.S. Ha, H.S. Kim, Y.G. Jeong, S.G. Han, H.K. Kim
Ji-Young Jeong, Sa-Woong Kim, Jun-Sung Chang, Tae-Hwan Kwon, Duck-Hoi Kim
Marianne Richou, Mehdi Firdaouss, Joris Fellinger, Thierry BAFFIE, Bernd Böswirth, Mariano Di Bartolomeo, Diogo Dias Alexio, Daniel Dickes, Daniel Dorow-Gerspach, Martin Draksler, et al.
The present study describes the ongoing developments for plasma-facing components (PFCs) intended for future fusion devices, requiring PFCs to handle high heat loads of at least 10 MW/m² in the divertor region. tungsten (W) is chosen as the armor material due to its resistance to plasma-wall interactions and high heat fluxes (HHF). Significant efforts are underway on various fusion devices (WEST, EAST, KSTAR…) to test and optimize water-cooled W-based PFCs. For W7-X, new PFCs are being developed, using W armor material placed at the divertor target region. The design aims to simplify manufacturing, inspection, and installation processes compared to the currently operated actively cooled carbon fiber composite based divertor, while meeting thermal and mechanical constraints. The developed concept and related manufacturing choices are promising since, we show that it is able to sustain the required heat loads. Also, the JT-60SA tokamak plans a transition to metallic PFCs after 2029. The W monoblock concept, successfully used in WEST, EAST and KSTAR and planned for ITER, is being adapted for JT-60SA, and is currently the reference concept. However, some advanced designs exploring cost-saving measures and enhanced heat transfer capabilities are also under development. Manufacturing processes, such as laser powder bed fusion (LPBF) and hot isostatic pressing (HIP), are optimized to ensure reliable cooling channels for the development of the proposed enhanced designs. For these developments, thermo-hydraulic analyses and HHF tests have shown promising results, with surface temperatures being in agreement with models and material temperature limits. These developments lead to show a successful use of combined advanced manufacturing processes (additive manufacturing…), design hypotheses and adapted modelling tools to propose components with relevant performances.
Changle Liu, Jiahui Pan, Xiaoyun Ding, Lei Li, Lei Cao, Damao Yao
Milan Zmitko, Yves Poitevin, Laurent Forest, Antonella Li Puma, Rémi Boullon, Jérôme Tosi, Denis Sornin, Philippe Emonot, Emmanuel Rigal, Laurence Cogneau, et al.
S. Julià, F. Albajar, M. Jimenez, T. Cicero, P. Estebanez, R. Morón Ballester, I. Eletxigerra, A. San Vicente, C. Gómez, E. Carbonell, et al.
The electron cyclotron (EC) upper launcher (UL) and ex-vessel waveguide (EW) are integral components of the ITER EC system (Sanchez et al 2025 Qualification of the European gyrotrons and power supplies of the ECH and CD system of ITER IAEA FEC ), which plays a key role in achieving and maintaining the high temperatures required for nuclear fusion within the tokamak, providing plasma heating and current drive, and magnetohydrodynamics stabilization, which are crucial for ITER’s operational success. The fusion for energy contribution to the current ITER baseline configuration comprises four ULs and five sets of EWs. The UL and EW systems, currently in its final design phase under the frame of a procurement agreement between Europe and ITER organization, have released the final design documentation package, which is here described, as input to the final design gate review that took place in July 2025. The full set of documents released is composed by: (1) a design description (DD) dossier composed by a UL & EW DD document and supported by a set of 3D models, drawings and diagrams; (2) a design justification dossier encompassing a set of end-to-end engineering analyses verifying the functionality and structural integrity of the components according to the applicable requirements, and the RAMI analysis of both systems; and (3) a product lifecycle dossier covering documentation for subsequent phases including manufacturing, assembly, installation, commissioning, operations, maintenance and decommissioning. Furthermore, the validation of the design is supported by currently on-going activities of prototyping and testing of the most critical components.
Zhihong Liu, Zhiyong Wang, Jianguo Ma, Huapeng Wu, Nengtao Zhou, Wangqi Shi, Tao Zhu, Yudong Su, Jiefeng Wu
Oxide dispersion-strengthened (ODS) steel is a promising structural material in fusion reactors, nanoparticles in matrix pin grain boundaries, and capture point defects and they delay the performance degradation caused by creep and irradiation. In this paper, RAFM steel with Y 2 O 3 was formed by selective laser melting (SLM) and the microstructural evolution under creep and irradiation were closely studied. SLM-formed ODS-RAFM steel had a longer high-temperature creep life than RAFM steel at the same stress and temperature, the fined grains were closely related to recrystallisation and pinning of oxide. The ODS-RAFM steel exhibited great irradiation stability, the lath width slightly increased, and the hardening rate was smaller than RAFM steel. During the irradiation process, continuously precipitated new MX carbides and Y–Ta–O oxide occurred in ODS-RAFM steel, which resisted the growth of grains. The above research results explained the microstructural evolution of RAFM steel and the change of performance under high temperature and an irradiation environment of fusion reactors, which verified the significant role of Y 2 O 3 nanoparticles in maintaining high temperature stability and irradiation resistance, hence providing a strong theoretical basis for the ODS-RAFM steel of the future to be fully used as the blanket structural material of nuclear fusion reactors.
G.H. Kim, Y.J. Jung, M.S. Ha, H.S. Kim, Y.G. Jeong, C.K. Park, H.K. Kim, Yu. Utin, S. Heggemann, T. Hübing
J. Willis, K. Chandrasekhar, P. Cheema, J. England, V. Godhani, E. Guise, S.M. Levine, R. Pocock, A. Scott, A. Shone, et al.
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.
Alexander Feichtmayer, Robert Lürbke, Alexander von Müller, Till Höschen, Andrei Galatanu, Max Olenberg, Rudolf Neu
G.V. Centomani, A. Iaboni, D. Avella, A. Buonocore, L. Barone, L. Cederle, M. De Bastiani, R. Durastanti, G. Ferrero, R. Forte, et al.
Emma Piozin, Mathieu Monteremand, Julien Cigna, Paolo Rossi, Jeong-Ha You
Chengwen Li, Zhihong Liu, Rui wang, Haibiao Ji, Jianguo Ma, Zhiyong Wang, Ren Chen, Aiming Liu, Shijun Qin, Huapeng Wu
Hisashi Serizawa, Hirotatsu Kishimoto, Hiroaki Mori, Takuya Ogura, Taichiro Kato, Yoshiaki Morisada, Hidetoshi Fujii, Naofumi Nakazato, Motoki Nakajima, Takashi Nozawa, et al.
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