Recent Publications

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-Level Tritium Transport Modeling for the Outer Fuel Cycle of the SCYLLA Breeding Blanket and Relevance for UNITY-2

Luigi Candido, Luke Taylor-King, Colin Baus, Christian Day, Italo Godoy-Morison, John McGrady, Minoru Jimma, Kyosuke Namba, Jonas Schwenzer, Tim Teichmann, et al.

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.

Experimental study of tritium release from neutron-irradiated lithium metatitanate using the thermal desorption method

Yevgen Chikhray, Saulet Askerbekov, Timur Kulsartov, Asset Shaimerdenov, Magzhan Aitkulov, Zhanar Bugybay, Assyl Akhanov

An Investigation of Surrogate-Assisted, Gradient-Based Methods for Stellarator Breeder Geometry Optimization

Connor A. Moreno, Paul P. H. Wilson

Neutron measurements with VERDI detectors in the long term irradiation station at JET tokamak during Deuterium-Tritium operation

Ion Evangelos Evangelos Stamatelatos, Marilia Savva, Theodora Vasilopoulou, K Mergia, S Messoloras, Steven Bradnam, Chantal Shand, Lee William Packer, Zamir Ghani, Paola Batistoni, et al.

The VERDI detector is a passive neutron detector using the multi-foil activation technique to measure neutron fluence in extreme fusion conditions. It features a low-activation capsule that contains a defined concentration of metallic elements. The robustness of the detector is ensured by the capsule material, while neutron detection relies on activation of the metallic elements. Post-irradiation gamma spectrometry combined with computational unfolding enables reconstruction of the neutron fluence and energy spectrum. This study reviews the development of the VERDI detector and reports the first deployment of VERDI detectors in an operational Deuterium-Tritium (D-T) fusion environment in the Long-Term Irradiation Station (LTIS) of JET tokamak during the DTE2 campaign, featuring a neutron spectrum representative of future fusion devices such as ITER and DEMO. The results provide benchmarks for validating neutronic simulations and nuclear data, supporting the design optimization and safety assessment of future fusion power plants.

Performance evaluation of CAD-to-MCNP conversion tools in complex fusion systems: A case study of VUV diagnostics in JT-60SA

S. Akbas, B. Bienkowska, A. Cubi, N. Hajnal, M. Cavinato, M. Di Giacomo, S. Soare, A. Belpane, M. Valisa, A. Fassina, et al.

Development of a phenomenological model of beryllide corrosion in water vapor

Zhanna Zaurbekova, Timur Kulsartov, Asset Shaimerdenov, Saulet Askerbekov, Inesh Kenzhina

Stress corrosion mechanism of CLF-1 RAFM steel in flowing Pb-17Li at 450℃: synergistic effects of tensile stress and microstructural evolution

Zhenchao Sun, Yu Guo, Xiujie Zhang, Teng Zhang, Wei Qian, Yao Zhao, Lei Wang, Xinting Lv, Yiming Wang, Zhengdong Li

The compatibility of RAFM steel with liquid Pb-17Li is a critical concern for its application as a structural material in liquid metal blankets of fusion reactors. This study investigated the stress corrosion mechanism of CLF-1 RAFM steel under a tensile stress of 250 MPa for 5000 hours in flowing Pb-17Li at 450 °C. The results indicate that stress-induced strain disrupts the continuity of the protective oxide layer, shortening the incubation period and increasing corrosion initiation sites, thereby intensifying the corrosion severity and susceptibility. The corrosion morphology is jointly influenced by the flow velocity of Pb-17Li and the steel’s microstructure. At high flow velocities, the martensitic lath structure is eroded, resulting in an etched surface. Conversely, lower velocities preserve the more corrosion-resistant laths, producing lamellar corrosion structures. Furthermore, shear stress promotes microstructural coarsening, which eliminates certain grain and sub-grain boundaries. Although this locally improves corrosion resistance, the resulting long and straight grain boundaries instead facilitates the penetration of liquid metal, thereby accelerating corrosion failure. These findings clarify the stress corrosion mechanism of RAFM steel in Pb-17Li and provide new insights for the modeling of liquid Pb-17Li corrosion and design of liquid metal blankets.

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

Numerical simulation of tritium extraction from liquid PbLi using multi-stage gas-liquid contactors

Kai Huang, Peng Ran, Shixin Ruan, Yongpeng Lyu, Teng Wang, Bin Zhang, Zhaoting Huang

Analysis of neutron single event effect experiment on electronics during JET DTE-3 campaigns

Martin Felix Dentan, Soilihi Moindjie, Matteo Cecchetto, Jean-Luc Autran, Ruben Garcia Alia, Richard Naish, John Waterhouse, Alan R Horton, Xavier L Litaudon, Fernanda Rimini, et al.

We measured SEEs in 40 nm and 65 nm SRAMs exposed to the D–T fusion neutron environment of the JET tokamak during the DTE-3 campaign (September–October 2023), while also characterizing the neutron spectrum seen by the devices. Using consolidated JET neutron production data, the measured SEE rates show good agreement with predictions based on the measured spectra and independently determined SRAM sensitivity parameters from thermal and monoenergetic neutron tests. These results extend the validation of our spectrum-based SEE prediction methodology, previously demonstrated in a D–D fusion neutron environment in 2021, to a D–T fusion neutron field. The novelty of this work is therefore the first experimental validation of this complete prediction approach in a D–T fusion environment, using real-time neutron spectrometry and independent device sensitivity data, rather than a new SEE model. The experiments also show that a local B₄C shield reduces the bit-flip rate by absorbing thermal neutrons, and we discuss the conditions under which such shielding can improve electronics reliability in tokamak and accelerator neutron environments.

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.

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

Hydrogen isotope permeation in liquid Li17Pb83 supported by pure V and V alloy substrates

Yuya Ando, Yuto Murata, Takuya Nagasaka, Teppei Otsuka

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

Neutronics Analysis of a Water-Cooled Blanket with a Be 12 Ti Reflector to Enhance Tritium Breeding

Seonghee Hong, Sungjin Kwon

Unstructured mesh-based global variance reduction method for deep-penetration fusion neutronics calculations in RMC

Jie Li, Pengfei Shen, Kok Yue Chan, Yongpeng Kong, Qing Li, Kan Wang