Recent Publications

Observations for a Fusion Tritium Economy

Mohamad Abdallah, Bjorn Cole, Daniel Clark, Jonas Kessing, Felipe Novais, Jonathan Naish, Paul Humrickhouse

BLUEMIRA: a modular, open-source framework for designing tokamak fusion reactors

Matti Coleman, James Cook, Fabrizio Franza, Ivan Maione, Simon Mcintosh, Hudson Baker, Alexander Ian Blair, Shail Desai, Oliver Funk, Georgina Graham, et al.

As efforts across the world to deliver fusion power enter their respective conceptual design stages, many begin to encounter the difficulties of generating complete designs, evaluating them, performing trade-off studies, and comparing different alternatives. Many organisations successfully use fusion reactor systems codes with low-fidelity formulations of the critical physics and engineering aspects to inform the initial design stage and steer them towards a region of the design space they consider attractive, but struggle to explore the design space in more depth. This work is an overview of the open-source tokamak reactor design framework, BLUEMIRA, built from the BLUEPRINT [1, 2] and MIRA [3, 4] codes, which goes beyond the “classical” systems code paradigm and enables the user to parameterise their reactor designs at higher fidelity. Reactor design workflows ranging from 0-D “radial build” design, equilibrium design, to 3-D CAD and automated neutronics analyses can be run within minutes. We present the functionality implemented in the BLUEMIRA code and demonstrate its application to the conceptual design of a conventional aspect ratio fusion reactor, performing an indicative design study over aspect ratio and number of toroidal field coils.

System-analysis-based investigation of geometric configuration effects in an in-vacuum-vessel loss-of-coolant accident for the CFETR vacuum vessel

Hongfei Zhu, Yangui Li, Yuliang Fang, Xiao Yan, Ming Ding, Xiaxin Cao

Adaptation of Nuclear Fission Cost Estimation Tools for Fusion Energy Systems

Matthew J. Nyberg, Virginia Lilly, Jia Zhou, Nicolas Stauff, Craig M. Jacobson, Ben A. Lindley

Preliminary Thermal-hydraulic Analysis of Subcritical Fission Zone for Conceptual Test Blanket Module Driven by Fusion Neutrons

Francesco Devia, Fabio Panza, Guglielmo Lomonaco, Seyed Kamal Mousavibalgehshiri

Further development of JADE as a tool for validation and verification of nuclear data libraries and particle transport codes

Alex Valentine, Steven Bradnam, Jude Moorehead, Davide Laghi, Alberto Bittesnich, Marta Campos Fornes, Matteo Di Giacomo, Marco Fabbri, Allan Harte

Several international leading efforts are focused on the improvement of evaluations for nuclear data. This data captures all of the physics of nuclear interactions for which the applications are far ranging from medical applications to criticality simulations. Nuclear analysis in the field of fusion, commonly referred to as neutronics, is one such application, typically employing Monte Carlo methods that rely on continuous-energy, pointwise cross-section data. Historically, the most widely adopted and complete radiation transport code in this field is MCNP, however in recent years there has been an evident trend in exploration of alternatives, most notably OpenMC. Nuclear analysis is a critical driver of the design, operation, safety and decommissioning of future fusion power plants stressing the importance of confidence in the transport codes and underlying nuclear data used in prediction of the radiation environment. Here we present recent developments to the JADE tool - a framework originally conceived for verification and validation of nuclear data within MCNP, now extended to performing cross-code comparisons. In particular, the developments required for the complete integration of OpenMC are detailed. A demonstration of this capability is presented for several of the currently implemented computational and experimental OpenMC benchmarks, with results compared against MCNP with the FENDL-3.2c and ENDF/B-VIII.0 nuclear data libraries.

Thermal-hydraulic analysis of the HTS part of the hybrid Central Solenoid designed for the next generation experimental fusion device

Monika Lewandowska, Aleksandra Dembkowska, Rafał Ortwein, Gianluca De Marzi

FED Thea energy special edition editorial

B. Geiger, S. Henneberg

Application of the RCC-MRx entrance key to the structural design of HCCP TBM-set components

Seong Dae Park, Suk-Kwon Kim, Jae-Sung Yoon, Hyoseong Gwon

The Helios cryostat, maintenance schemes, and cryogenic production system

J. Wasserman, W. Harris, N. de Boer, K. Brodek, J. Jeon, S. Pasmann, C.P.S. Swanson, S. Walsh, B. Berzin, D.A. Gates

A structural beam-based approach for preliminary structural sizing of toroidal coil system: A supplementary module to MADE algorithm

Lorenzo Giannini, Laura Pittaluga, Giuseppe Ruta

Thermal-hydraulic analysis of energy storage and intermediate heat transfer systems for tokamak fusion reactors

Francesco Colliva, Cristiano Ciurluini, Bruno Gonfiotti, Amelia Tincani, Fabio Giannetti

First Plasma Commissioning and Operational Highlights from India's First Spherical Tokamak at IPR

Kishore Mishra, Aditya Verma, N. Mansoori, Saurabh Verma, Y. Paravastu, M. S. Khan, Arvind Kumar, S. G. Thatipamula, Vishal Verma, M Sheetal, et al.

A compact Spherical Tokamak(ST) is commissioned at Institute for Plasma Research (IPR) to explore low aspect ratio tokamak physics and technologies that complement to the existing high aspect ratio tokamaks namely ADITYA-U and SST-1 by enabling studies on non-inductive startup, current drive in over dense plasmas, and shaped plasma physics on a low cost platform. The device, India's first spherical tokamak has completed major mechanical, magnetic, and electrical integration, and the coil system has been successfully tested with series of integrated commissioning. First plasma experiments have been carried out with a modest Ohmic system assisted by a 2.45GHz microwave system, supported by a centralized control and data acquisition system. An initial diagnostic set comprising visible imaging, spectroscopy, magnetics, and radiation monitors required for machine operation has been installed. This paper presents the integrated commissioning experiences and first plasma experiments of the newly installed machine.

Compact Experimental Negative TriAngUlarity Reactor (CENTAUR): A design study for a compact, affordable breakeven tokamak

The CENTAUR Collaboration, Samuel W. Freiberger, Evan Bursch, Javier Chiriboga, Hiro J. Farre-Kaga, Eliot Felske, Sophia Guizzo, John Labbate, Shreyas Seethalla, Frederick Sheehan, et al.

This work presents the compact experimental negative triangularity reactor (CENTAUR), a low overnight cost, high-field tokamak, breakeven reactor design, achieving a predicted total fusion power of 40MW and scientific energy gain of 1.3. Ballooning stability calculations confirm that the device's pedestal is within the first stability regime, which is consistent with the expected ELM-free operation associated with negative triangularity (NT) plasmas. The geometry of the NT divertor allows for high fraction of radiated power (13.5$\%$) between the separatrix and plasma facing components. Heat transport modeling based on simulations of the edge region show heat loads into plasma facing components well below material limits. The magnet system employs rare-earth barium copper oxide (REBCO) high-temperature superconductors in 18 toroidal field coils, an hourglass-shaped central solenoid, and six poloidal field coils to support high-field ($B_0=10.9$ T) plasma confinement, shaping, and current drive. Neutronics analysis shows that a 12 cm $B_4C$ shield keeps superconducting magnet heating below the 33~K quench limit during 10 s, 40 MW DT pulses. With this shielding, the modeled fluence indicates HTS components can survive more than ten times the 3000-pulse design lifetime. Iteration of economic analysis in tandem with the technical design process allows CENTAUR to achieve its overnight cost goal of $\$$2B determined using a custom costing model that predicts a total overnight cost of $1.6$B$\pm0.2$B.

Toward the design validation of water-cooled ceramic Breeder Test Blanket Module in physical mock-up testing

Wenhai Guan, Tsutomu Osaka, Takanori Hirose, Hideyuki Ide, Takehiro Kouno, Osamu Sawahata, Yoshinori Kawamura, Hiroyasu Tanigawa

A large-scale heat flux testing methodology for the water‑cooled ceramic breeder (WCCB) Test Blanket Module (TBM) submodule, the leading blanket concept for Japan’s DEMO reactor, was developed and demonstrated using a newly constructed electron‑beam (EB) heat load facility. To reproduce ITER‑relevant thermal conditions, dedicated calibration techniques were established to determine the net heat flux intensity, EB beam profile, and non‑destructive surface temperature measurement capability. A calorimeter composed of seven F82H heat‑receiving elements was designed to quantify the net heat flux reaching the mock‑up surface. Using this method, the nominal ITER TBM heat flux of 0.30 MW/m² was successfully achieved with a spatial variation within ±0.02 MW/m², establishing optimal EB parameters of Ic = 550 mA and Ub = 30 kV. The EB beam profile, evaluated using a newly developed Faraday cup with adjustable shielding, exhibited a Gaussian distribution with a diameter of 27.6 mm. A non‑contact infrared (IR) temperature measurement technique was developed using a dedicated calibration specimen, achieving agreement with thermocouple measurements within 6.6% up to 550°C. Based on these calibrations, a full‑scale physical mock‑up of the WCCB TBM submodule was irradiated at 0.30 MW/m² for 60 s under PWR‑like coolant conditions (280°C, 15.5 MPa). The measured temperature evolution showed good agreement with transient thermal analysis in the initial heating phase, while steady‑state deviations were attributed to EB‑induced surface emissivity changes and oxidation. Post‑test examinations based on ITER divertor acceptance criteria revealed no melting, coolant leakage, weld detachment, or thermal hot spots. These results demonstrate the validity of both the WCCB TBM design and the developed heat flux testing techniques, establishing a robust foundation for upcoming qualification tests toward ITER installation.

On the feasibility of model-based feedback control of vertical instability growth rate using out-vessel coils in ARC-like scenarios

Arunav Kumar, Cesar Clauser, Theodore Golfinopoulos, Jon C. Hillesheim

In this work, we propose a model-based feedback controller that regulates the vertical instability growth rate ($γ_{gr}$) of a high-elongation, double-null tokamak directly, using only out-vessel poloidal field (PF) coils. High elongation raises the achievable plasma current and fusion performance but makes the plasma vertically unstable, and in a fusion power plant the in-vessel coils that present devices rely on for stabilization may be absent, leaving only distant out-vessel circuits. The controller couples a machine learning surrogate of non-rigid, profile agnostic vertical instability metric to a constrained quadratic program: the surrogate supplies real-time $γ_{gr}$ estimates and, via automatic differentiation, the actuator sensitivities, while the program allocates coil voltages to track a target growth rate, maintain double-null divertor balance, and respect electromechanical limits. We tested this method on the ARC~V3A power plant design configuration across 24 closed-loop simulations spanning equilibrium variations, actuator degradations, and transient disturbances. We achieved full or marginal success in 83\% of these cases (full in 50\%, marginal in a further 33\%) and lose control in the remaining 17\%; the failures map the boundary of out-vessel controllability (occurring at the highest growth rates) and under actuator limits. The controller does not regulate boundary shape explicitly: separatrix geometry follows indirectly from growth rate and flux balance control and would require a separate shape control layer for sustained scenario evolution.

Tritium Management in an All-Carbon First Wall Tokamak Reactor

James Davis, Sellathurai Suppiah, Ezekial Unterberg, Donald Ryland, Peter Stangeby

Neutronic Analyses for the Design of the Chamber of Xcimer’s Athena Fusion Pilot Plant

Francisco Ogando, Robert Earley, Wayne Meier, Rene Raffray, Patrick Sauvan, Jaime Marian, Susana Reyes, Conner Galloway

Tuning of generalized k-omega turbulence model for prediction of heat transfer and pressure drop for helium cooled demo FW

Christine Klein, Frederik Arbeiter

Preliminary design of DTT vacuum pumping system

Mariagrazia Romano, Michele Fincato, Paolo Innocente, Christian Avanzato, Luca Balbinot, Andrea Belpane