Recent Publications

Failure Analysis and Engineering Optimization of an In-vessel Fiber-optic Current Sensor Routing Structure in the EXL-50U Spherical Tokamak

Jia Li, Dong Guo, Liu Shuo, Lombroni Riccardo, Zhixin Wang, Renyi Tao, Yumin Wang, Yuejiang Shi

Plasma current measurement is essential for tokamak equilibrium, control, and machine protection. Fiber-optic current sensors (FOCS) provide electrical isolation and immunity to electromagnetic interference, while their metallic in-vessel protective routes remain subject to electromagnetic loads and vacuum-boundary constraints. An annealed-copper FOCS protective tube in EXL-50U developed a vacuum leak, and inspection found localized flattening, rubbing, blackened surfaces, and fracture-like damage near center-column supports. Possible failure mechanisms were assessed using the assembly geometry, inspection evidence, electromagnetic estimates, and an idealized ring-compression reference. A representative normal toroidal-field (TF) ramp gives a loop current scale of approximately 1.4 kA and a distributed line load of 3.2 N/mm. A representative 500 kA vertical displacement event yields a toroidal-flux change of about 0.0639 Wb over a 0.1 ms output interval, corresponding to a finite-difference voltage scale of 639 V. An ideal impulse calculation gives currents of 1.6–12.8 kA for the assumed coupling and inductance range. The estimates show the scale of possible electromagnetic loading, but do not determine the local flattening threshold or the exact cause of failure. The observations suggest that tube motion against relatively rigid surrounding structures contributed to local deformation and possible electrical or thermal damage. The revised design uses a 316L stainless-steel tube, additional supports, polyimide insulation at support interfaces, and improved seals. We also discuss the design limitations and a ceramic electrical break as an alternative.

Neutron wall loading in stellarator optimization

Enrique Miralles-Dolz, R. Michael Churchill, Dario Giovanni Panici, Jacob Schwartz, Tim Bohm, Connor Moreno, Paul P H Wilson

Stellarator neutronics is inherently three-dimensional, since non-axisymmetric equilibria produce spatially varying neutron wall loading (NWL). Localized NWL peaks affect first-wall lifetime, shielding requirements, magnet protection, maintenance planning, and ultimately plant availability, but resolving these peaks with high-fidelity Monte Carlo transport remains too expensive for direct use in stellarator optimization. This work integrates a deterministic line-of-sight NWL model into the DESC stellarator optimization package so that NWL can be evaluated as a fast, differentiable engineering metric during stage-I equilibrium optimization. The model treats the deuterium-tritium plasma as a volumetric neutron source, evaluates the Bosch-Hale fusion reactivity from prescribed flux-surface profiles, and maps source contributions to first-wall points through a geometry-dependent kernel. The resulting local NWL map can be used directly in optimization objectives or constraints. The DESC implementation is validated against ParaStell/OpenMC Monte Carlo calculations for circular axisymmetric, ITER-like axisymmetric, and non-axisymmetric configurations; the surface-averaged NWL agrees within about 1.5% across these cases and the deterministic maps reproduce the dominant spatial loading structures while tending to conservatively overestimate the Monte Carlo wall loading. In proof-of-concept optimizations, current-profile variation at fixed plasma boundary reduced peak NWL and the peaking factor by 13.84% and 13.29%, respectively, while quasi-symmetry optimization through plasma boundary shaping reduced peak and mean NWL by 17.22% and 12.97%. These results demonstrate that differentiable NWL objectives and constraints can reveal useful engineering design directions unavailable to conventional plasma physics objectives.

Analytical pre-screening and PROCESS initialization for spherical tokamak design studies

Oreoluwa Ajayi, Hossam A. Gabbar

Analysis of supercritical CO2 brayton cycle system for laser inertial confinement fusion concept reactors

Bingqian Zhang, Xinze Li, Ronghua Chen, Kui Zhang, Wenxi Tian

How to make nuclear fusion “XR-ready”: A role-based CAD-to-XR framework

Sara Buonocore, Andrea Fimiani, Francesco Marino, Venero Russo, Andrea Zoppoli, Raffaele De Amicis, Giuseppe Di Gironimo

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

Designing a buildable optimised stellarator to confine electron–positron plasmas

Pedro Francisco Gil, Jason Smoniewski, Paul Huslage, Rogerio Jorge, Timo Thun, Elisa Buglione-Ceresa, Tristan Schuler, Stefan Fingl, Grégoire-Hubert Ducas, Eve Virginia Stenson

In this paper, the design of the plasma equilibrium and superconducting coils for the electrons and positrons in an optimised stellarator electrons and positrons in an optimized stellarator (EPOS), experiment is presented. With newly developed stellarator optimisation tools, including single-stage and stochastic optimisation, as well as high-temperature superconductor strain, this work demonstrates that it is possible to achieve key metrics for the buildability and confinement properties of the device. In particular, satisfactory quality of quasisymmetry and stellarator robustness is designed, and engineering requirements are met for eight different candidates. A feasibility study is presented that optimises multiple candidates for different plasma major radii and coil currents, as well as the best EPOS candidate to date, named C4_R19.

Progress in reactor-core fusion technologies presented at the 30th IAEA Fusion Energy Conference 2025

Jiming Chen, Jianbao Wang, Qixiang Cao, Shen Qu, Zongyu Yang

This paper provides an overview of the recent progress in reactor-core fusion technologies within the vacuum vessel presented at the 30th International Atomic Energy Agency (IAEA) Fusion Energy Conference held in October 2025 in Chengdu, China, including plasma control, neutronics and nuclear data, tritium technologies, design and development of in-vessel components, related materials and intense neutron sources for material irradiation. Plasma control focusing on achievements for ITER and major fusion research devices are covered and the application of digital twin and Artificial Intelligence are highlighted. Research and Development (R&D) results from the International Thermonuclear Experimental Reactor (ITER) and current devices to future Demonstration Fusion Power Plant (DEMO) reactors are summarized, particularly new findings, data validation, design and analysis tool improvement, properties assessment and technical solutions for critical issues.

Whole Plant Design, Integration, and Build

J. Morris

Neutron-induced air activation in the ITER HRNS diagnostic enclosure: Quantitative assessment and radiological implications

Anna Wójcik-Gargula, Grzegorz Tracz, Rojin Mehrara, Urszula Wiącek, Giovanni Mariano, Bruno Coriton, Ryszard Kantor, Dariusz Morawski, Marek Scholz

Engineering Design Frameworks for Commercial Fusion Energy Systems: An Assessment of Technological Architectures, Materials Qualification, and Tritium Fuel Cycle Integration

Alberto Boretti

Development and preliminary validation of a Nek5000-based magnetohydrodynamics solver for fusion blanket relevant liquid metal and molten salt flows

Filipe L. Brandao, Arpan Sircar, Marco Delchini, Vittorio Badalassi

Thermodynamic and exergy analysis of a multi-source TES-coupled Rankine cycle for EU DEMO with DCLL blanket

Yuan Sun, Minyang Gui, Xiaodong Liu, ZhaoJun Li, Haibo Jia, Ruifeng Tian, Sen Duan, Sichao Tan

Comparison of the pressure drop and flow characteristics between two designs of a liquid metal blanket supply system

Yuchen Jiang, Yuqiao (Joy) Fan, Sergey Smolentsev

Systems analysis of reactor-scale stellarators with optimised plasma–coil spacing

Jedrzej Walkowiak, Felix Warmer, Sehyun Kwak, Andrea Pavone

This study evaluates the feasibility of a stellarator power plant using a new Stable Quasi-Isodynamic Designs (SQuID) configuration within the PROCESS systems code. Our results indicate that for a stellarator constructed with high-temperature superconducting (HTS) technology, a neutron wall load is a primary restriction for size reduction. With an optimistic assumption of 1.5 MW m−2 neutron wall load limit, a major radius of 16.5 m is achievable for 1GW of plant net electric power. We achieve the reduction in size by changing the coil radius in relation to major radius. It is shown that right preselection of the minimal plasma-coil distance at the stage of coil set optimization allows to find optimal solution in machine optimization. Analysis of the proposed coil set suggests that both HTS and low-temperature superconducting (LTS) technologies are viable magnet options. The PROCESS cost model shows that HTS becomes economically advantageous when its material cost is less than five times that of LTS, assuming comparable manufacturing complexity and risk.

Indirect-Drive Fusion Target Design for Commercial Fusion Energy

C. R. Weber, A. L. Kritcher, S. Bhandarkar, T. Briggs, T. Chapman, T. M. Fears, B. A. Hammel, D. D. -M. Ho, O. Hurricane, B. Kozioziemski, et al.

This paper presents the physics basis for commercially relevant laser indirect-drive (LID) (radiation-driven) inertial fusion energy (IFE) using a 10 MJ laser driver. To date, this approach, proven at the National Ignition Facility (NIF), remains the first and only controlled fusion method to demonstrate the key physics required for fusion energy production, including a self-sustained burning plasma, substantially de-risking the path to commercial fusion energy. Building directly on these results, we present scaled designs to larger target sizes and fusion gains relevant for commercial power generation ($G\sim 26$--$43$). The designs remain close to experimentally demonstrated ignition physics, modifying target components to improve scalability, manufacturability, and cost-effectiveness for fusion energy applications while preserving ignition-relevant implosion physics and fusion power plant compatibility. The baseline platform retains a high-density carbon ablator and clean cryogenic DT fuel layering while extending ignition platforms to substantially larger fuel masses (exceeding 10 times that of current ignition experiments at the NIF) and higher burn fractions ($\sim$40\%) with total areal densities at stagnation of $\sim$3~g/cm$^2$. Benchmarked simulations anchored to NIF ignition experiments (using HYDRA and LASNEX) predict that these designs achieve robust ignition and propagating burn at substantially higher fusion yields (265--427~MJ) with significant ignition margin (2--4$\times$ relative to NIF) against hydrodynamic instabilities and representative power-plant non-idealities, including low-mode asymmetry, polycrystalline DT ice roughness, HDC ablator voids, and target-support and fill-hole perturbations. We also show that implosion symmetry and laser-plasma interactions (LPI) can be controlled with our novel multi-beam configuration using thousands of laser beam-lines.

Extensions to the ITER actuator management architecture for dynamic actuator allocation and pellet system reconfiguration

O. Kudláček, S. Stanek, W. Treutterer, I. Gomez Ortiz, P. Lang, R. Nouailletas, L. Pangione, T. Ravensbergen, M. Reich, J. Schmalz, et al.

Spatially resolved evaluation of primary displacement damage and activation in the CFETR first-wall components

Yingming Sang, Zeenat Ullah, Kuanhao Liu, Shehu Adam Ibrahim, Da Chen

WEST Operation - Reliability and availability of a long pulse fusion tokamak

Valérie Lamaison, Cyril Brun, Elodie Corbel, Annika Ekedahl, Laurent Gargiulo, Sebastien Hacquin, Michael Houry, Lionel Meunier, Philippe Moreau, Lionel Toulouse

Since 2016, the WEST tokamak has demonstrated its capability to perform long plasma discharges approaching 1000 seconds in a fully metallic environment. It operates with a permanent magnetic field of up to 3.65T generated by 18 superconducting toroidal field coils cooled with helium at 1.8K produced by a cryogenic system. Since 2021, all plasma-facing components (PFCs), including the tungsten ITER-grade divertor, are actively cooled by pressurized water, making WEST representative of future superconducting fusion devices. Between 2022 and 2024, experimental campaigns achieved significant improvements in performance. The number of long-duration discharges (>100s) increased threefold thanks to the non-inductive current drive from the Lower Hybrid Current Drive (LHCD) system, culminating in a world record plasma duration of 22 minutes with 2.6 GJ injected energy. Total plasma time exceeded five hours per year, with over 70% of successful pulses. These results are enabled thanks to the availability of the WEST machine and all subsystems, higher than 70%. The study of downtimes recorded during the last three years experimental campaigns, shows four main elements/systems impacting WEST operation: the water and air leaks in vacuum vessel, the poloidal field system, the CODAC (Control, Data Access and Communication) system and the cryogenic system. Key lessons for future fusion devices to achieve plasma long pulses include a high availability of the machine based on a targeted maintenance plan to maximize reliability of all sub-systems and an effective responsiveness in incident diagnosis and repair, particularly in water leak detection on actively cooled Plasma-Facing Components (PFC).

Improved n=1 empirical error field penetration threshold scaling with Ohmic and L-mode conventional tokamak plasma discharges

Evan Maxwell Bursch, Jong-Kyu Park, Nikolas C. Logan, Feiyue Mao, Nengchao Wang, Carl Friedrich Benedikt Zimmermann, Richard J Buttery, Carlos Paz-Soldan, Matthew Pharr, Lidia Piron, et al.

This paper presents an updated n=1 error field penetration threshold scaling, which increases fit quality compared to previous error field scaling laws, is produced from an expanded database, and exhibits reduced uncertainty in projections to future tokamaks. It improves confidence in tokamak engineering tolerances, which are a significant driver of cost and time constraints on device construction. We add J-TEXT data, new JET data, and create the scaling using only conventional tokamak Ohmic and L-mode experiments. Since H-mode plasmas are more resilient to error field penetration, this scaling predicts what is likely the most dangerous regime of error field penetration for new tokamak designs. These decisions improve confidence in the error field penetration threshold scaling and its application in the construction and design decisions of any future conventional tokamak or fusion pilot plant.