Recent Publications

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

Yuya Ando, Yuto Murata, Takuya Nagasaka, Teppei Otsuka

Tritium Management in an All-Carbon First Wall Tokamak Reactor

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

University of Toronto, Canadian Nuclear Laboratories, Oak Ridge National Laboratory

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

Xcimer Energy Corporation, Universidad Nacional de Educación a Distancia (UNED), Lawrence Livermore National Laboratory, University of California, University of California Los Angeles

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

Seonghee Hong, Sungjin Kwon

Korea Institute of Fusion Energy

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

Tsinghua University, Nuclear Power Institute of China, National University of Defense Technology

Mitigation of sawtooth transient heat flux in the MAST-U Super-X divertor with deuterium and nitrogen

Rory Scannell, Jack Flanagan, Zhouji Huang, James R Harrison, Kevin Verhaegh, Scott Alan Silburn, Peter Ryan, Stuart Scott Henderson, Nicola Lonigro, Sidney Leigh, et al.

Culham Centre for Fusion Energy, UKAEA, University of Liverpool, Eindhoven University of Technology, EURATOM/CCFE Fusion Association

The MAST Upgrade Super-X divertor protects plasma-facing components from heat fluxes during transient events. This paper examines heat loads from sawtooth crashes with energies ΔWsawtooth ≈ 2–9kJ and investigates the impact of deuterium and nitrogen gas pressure on their mitigation. In a D2 gas scan for sawteeth with energies up to 6kJ, the peak heat flux (q⊥,max) scales with the transient energy and inversely with the neutral pressure, with variations in ΔW/Pgas accounting for 70% of the observed variation in q⊥,max. Larger sawtooth transients (≈ 6–9 kJ) fall outside this trend and exhibit much higher heat fluxes. With D2 puffing in the divertor, the Super-X configuration exhibits substantially lower peak q⊥,max for comparable transient energy loads than observed in conventional divertor configurations. However, this reduction is broadly consistent with geometric expectations especially at higher pressures. Nitrogen gas seeding was applied to transients with energies below ≈ 6 kJ. Increasing the nitrogen pressure reduces q⊥,max, by 4.5MW/m2/Pa in Super-X while nitrogen seeding is found to have much less impact on transients in conventional divertor configuration. In the case of N2 seeding, the Super-X divertor outperforms conventional divertor configuration beyond geometric expectations. Direct Super-X divertor Te and ne measurements during transients were obtained from Thomson scattering. These indicate quiescent inter-transient temperatures of < 1 eV, corresponding to strong detachment. The temperatures rise to 6–8 eV during typical sawtooth transients and 10 eV where divertor heat loads exceed q⊥,max > 2MWm−2. The experimental data are compared to modelling results using a 1D exhaust code in the ReMKiT1D framework. The modelling highlights the importance of recycling on understanding the buffering of transients.

Development of a flow distributor for the actively pumped open-surface lithium loop

Steven Stemmley, Daniel O'Dea, Braden Moore, Riley Trendler, Quentin Mooney, Emily Greene, David N. Ruzic

University of Illinois Urbana-Champaign, Tokamak Energy Ltd.

Experimental study on dynamic evolution of liquid lithium–plasma interaction and thermal response based on 3d-printed finger-type capillary porous structures

Xichen Su, Rendeng Tang, Fujun Gou, jianjun Wei, Zongbiao Ye

Sichuan University

Thermomechanical analysis of the conceptual design of the ITER loss alpha monitor probe head

J. Hidalgo-Salaverri, J. Gonzalez-Martin, M. Kocan, J. Ayllon-Guerola, C. Navarro, E. Agus, G. Simic, M. Garcia-Munoz, J. Rueda-Rueda, R. Marques

University of Seville, Politecnico di Torino

Tungsten limiter start-up experiments on ASDEX Upgrade and WEST in different boronization states in support of ITER

Joerg Hobirk, Richard A Pitts, Pierre Manas, Clemente Angioni, Matthias Bernert, Dominik Brida, Guido Ciraolo, Laurent Colas, Corinne Desgranges, Ralph Dux, et al.

Max-Planck-Institut fuer Plasmaphysik, CEA, ITER Organization

Understanding the performance of limiter plasmas in the ITER start-up phase is important for the whole pulse and a possible challenge if performed on tungsten, as will now be the case following the switch to a W first wall in the new 2024 ITER Baseline. Experiments were performed on ASDEX Upgrade and WEST to characterise limiter plasmas using boronizations with different degree of boron surface coverage and toroidal asymmetries. Non-boronized start-up is shown to be slow and laborious, and, on both machines, was aborted in favour of a non-homogeneous boronization, performed also in support of the ITER re- baseline to study the impact of spatially non-uniform boron coatings. This allows normal start- up and short (few 100 ms, ITER will run ≈ 10s long limiter phases) limiter phases can be run without problems. Even with a full boronization, the limiter can de-condition and long limiter plasmas of several seconds suffer from high densities and radiation. Short limiter plasmas for plasma current ramp-up remain possible. The conditioning effect on limiter plasmas is documented, but also how the start-up is affected by an ageing boronization.

Precision Machining of M72 Inconel 718 Studs for ITER Vacuum Vessel Gravity Support

G.H. Kim, Y.J. Jung, M.S. Ha, H.S. Kim, Y.G. Jeong, S.G. Han, H.K. Kim

Mechanical characterization of fine-grain dispersion-strengthened tungsten as a plasma facing material

Gabriel D. Parker, Zachariah Koyn, Tim Graening, Hsin Wang, Xiang (Frank) Chen, Patrick E. Albert, Xiao-Ying Yu

Oak Ridge National Laboratory, Energy Driven Technologies LLC, The Pennsylvania State University

Mesh-based multiphysics coupling acceleration for fusion neutronics clustering for fusion blanket applications

Jin Whan Bae, Arpan Sircar, Katarzyna Borowiec, Vittorio Badalassi, Cami Collins

Oak Ridge National Laboratory

Accurate modeling of particle transport within fusion blankets is essential for predicting performance metrics such as heat deposition and the tritium breeding ratio (TBR). However, high-fidelity coupling of thermal fluids from computational fluid dynamics (CFD) to neutronics simulations often incurs significant computational costs due to the complexity of surface intersection calculations in Monte Carlo codes. This paper presents an accelerated multiphysics coupling method for neutronics that utilizes hierarchical agglomerative clustering to map complex material property distributions to a neutronics model. Implemented within the Fusion Reactor Design and Assessment (FREDA) framework, the method leverages existing Python packages to automate the creation of clustered geometries for OpenMC. The approach is demonstrated on a sector model of an ARC-class tokamak with an immersion molten salt blanket, and an simple geometry with varying isotopic concentrations. Results show that the clustering method significantly reduces computational burden without compromising fidelity, providing a foundation for agile iteration of neutronics simulations involving multiple coupled material properties.

Compatibility of Type 304H Stainless Steel in Static Molten FLiBe for Inertial Confinement Fusion Reactors: Role of Impurities and Redox Control

Rishi Pillai, Debashish Sur, Severine Cambier, Adam Willoughby, Tracie Lowe, Youyang Zhao, Amanda Bye, Susana Reyes

Oak Ridge National Laboratory, XCIMER Energy Corporation

Modeling of divertor heat flux limits and lithium vapor shielding in NSTX-U using UEDGE code

MD Shahinul Islam, Maxim V Umansky, Vlad Soukhanovskii

Lawrence Livermore National Laboratory

In this work, UEDGE simulations coupled self-consistently with a two-dimensional wall transport solver (Wall-Li) are used to investigate two closely related issues important for NSTX-U operation: (1) plasma-surface interactions of graphite plasma-facing components (PFCs) under increasing input power for standard and snowflake divertors and (2) lithium PFC vapor shielding and its dependence on upstream plasma conditions for a standard lower single null divertor with a 5~mm layer of lithium on the graphite tiles. Simulations with graphite target PFCs show a monotonic increase in surface temperature with increasing heat flux incident on the target plates. For a broad range of operating conditions, surface temperatures exceed the graphite sublimation temperature (>1200°C, when graphite starts massively sublimating) once peak heat fluxes exceed 7~MW/m2. This occurs despite enhanced carbon radiation. A snowflake-minus divertor configuration is analyzed to assess divertor heat handling on the secondary strike point (SP\#2) under conditions where both magnetic field-line incidence and cross-field transport are uncertain. Simulation results indicate that the heat flux to SP\#2 is dominated by enhanced CM transport rather than parallel conduction, leading to increased deposited heat flux with increasing CM strength and incidence angle. Lithium vapor shielding is investigated using a self-consistent model in which lithium sourcing depends on local plasma and surface conditions. The model demonstrates that lithium vapor shielding provides passive thermal cooling, maintaining surface temperatures below 700°C even as the core power increases. However, lithium accumulation upstream increases once surface temperatures exceed 600°C. Increasing the core density augments the drag force on impurity ions due to the parallel plasma flow in the scrape-off layer, confining lithium closer to the target and thereby extending the operational window for effective vapor shielding. Overall, these results highlight the necessity of (1) thick lithium layers to prevent strike-point depletion, (2) active cooling mechanisms (such as fast-flowing liquid lithium) to suppress excessive surface temperatures and evaporation, as lithium operation is constrained by surface temperature, and/or (3) deuterium gas puffing near the divertor to further increase SOL flow and density

Overview of the European Breeding Blanket programme for Helias 5-B: Dual Coolant Lead-Lithium design, integration and tools

Iole Palermo, Javier Alguacil, Gaetano Bongiovi, Juan Pablo Catalan, Ilenia Catanzaro, Iván Fernández-Berceruelo, Salvatore Giambrone, Guillermo Gómez Fonfría, Jose Ángel Noguerón, Vicente Manuel Queral Mas, et al.

CIEMAT, Universita degli Studi di Palermo, UNED

As part of EUROfusion’s mission to bring stellarators to technological maturity, the Stellarator Power Plant Studies (SPPS) WPPRD began in 2021 to develop a HELIAS-class power plant. Building on DEMO tokamak experience, European teams are designing a Dual Coolant Lead-Lithium (DCLL) breeding blanket (BB) for HELIAS. This concept uses liquid PbLi as breeder/coolant and decoupled helium cooling for the first wall (FW). Two key adaptations address HELIAS’s complex geometry: a detached FW using Capillary Porous System (CPS) technology, and a quasi-toroidal segmentation (QTS) with PbLi flow aligned to magnetic field lines. QTS reduces magnetohydrodynamic (MHD) pressure drop by up to two orders of magnitude, potentially eliminating electrical insulation needs. Remote handling (RH) is rethought for 3D stellarators, where traditional vertical-port extraction is impractical. Alternatives include enlarged fixed coils, movable coils for temporary large ports, and detachable vessel periods. The detached FW strategy shifts maintenance from large BB segments to smaller, easily replaceable FW panels, extending BB lifetime. CPS designs with Li in tungsten matrices were analysed thermally, hydraulically, and neutronically, showing potential to lower displacement-per-atom (dpa) damage while maintaining tritium breeding. Be-based moderators behind the FW CPS matrix improved TBR while reducing back-structure damage. To accelerate design and analysis, HeliasGeom and SHANE tools were developed to rapidly generate realistic 3D parametric geometries for CAD, neutronic, and thermal-hydraulic coupling. Preliminary 3D studies addressed TBR optimisation, shielding, MHD in non-uniform fields (via GridapMHD), and multi-scale thermal–mechanical assessments. These innovations collectively advance stellarator blanket technology, integration, and maintainability toward viable power plant concepts.

Multiscale assessment of tritium behavior in preliminary fusion pilot plant design using surrogate models in TMAP8

Lin Yang, Pierre-Clément Simon, Emre Yildirim, Jose Manuel Trueba Cutillas, Matthew Robinson, Masashi Shimada

Idaho National Laboratory, Tokamak Energy Ltd

The complexity and significance of multiscale phenomena in fusion energy systems make advanced modeling necessary for designing, optimizing, and safely deploying fusion plants. Tritium accountancy is one of those challenges for deuterium–tritium fusion systems. Its availability is constrained by its short half-life (12.33 years) and limited natural abundance, which require fusion plants to breed tritium onsite. Therefore, accurate tritium accountancy is essential for effective resource management, safety, and economics in fusion plants. Through the U.S. Department of Energy milestone program, Tokamak Energy Ltd. is developing a fusion pilot plant design and evaluating tritium retention and loss in key components and their effect on the fuel cycle. To rapidly explore design trade-offs and quantify design decisions on tritium management, this study presents a multiscale analysis to investigate tritium diffusion, trapping, and recovery in key plasma-facing components. To enhance computational efficiency, we integrate surrogate models at the component-level within a fuel cycle model at the system-level, enabling rapid evaluation of tritium recycling dynamics and inventory under various operational scenarios. The goal of this study is twofold: (1) demonstrate the feasibility of utilizing surrogate models to increase the accuracy of fuel cycle modeling, and (2) rapidly evaluate the performance of fusion technologies to accelerate design iterations. This multiscale model provides the tritium transport and retention behavior and supports the plasma-facing components design optimization in normal and bake-out operations. The work is implemented using the Tritium Migration Analysis Program, Version 8 (TMAP8), an open-source application for tritium transport analysis in fusion systems.

Scalable production of nuclear battery alpha emitters using fusion neutrons

J. F. Parisi

Nuclear batteries powered by alpha decay have been deployed successfully for over 60 years, on a worldwide $^{238}$Pu supply of kilograms per year. We show that the 14 MeV neutrons of a single deuterium-tritium fusion plant can produce alpha emitter battery fuels up to tons per year, in three classes: fuels with completely new production pathways ($^{236}$Pu, $^{227}$Ac, $^{210}$Pb), fuels previously proposed whose scarce feedstock the same pathways now breed at scale ($^{232}$U, $^{228}$Th), and the established $^{238}$Pu. OpenMC simulations of actinide channels in a tokamak blanket give, per GW yr of fusion: 11 to 57 kg of $^{236}$Pu, whose chain releases 18 GJ per gram over a century, ending at stable $^{208}$Pb, plus up to 5.2 t of co-product $^{238}$Pu; up to 1.4 t of $^{231}$Pa from thorium, and, from channel with $^{231}$Pa feedstock, up to $\sim$15 t of $^{232}$U or $\sim$122 kg of $^{210}$Pb, with $^{227}$Ac produced at 21 g/yr per tonne of $^{231}$Pa. Neutron capture also upgrades $^{241}$Am to a $^{242}$Cm/$^{242m}$Am/$^{241}$Am/$^{238}$Pu blend with up to 10 times higher power density. The same $^{236}$Pu and $^{232}$U also serve as proliferation safeguards: the $^{237}$Np, $^{232}$Th, and $^{231}$Pa channel products are self-protecting, the plutonium by $^{236}$Pu and $^{238}$Pu decay heat and the 2.6 MeV gammas from $^{208}$Tl content, and similarly the uranium from its $^{232}$U. Many of these fuels ($^{236}$Pu, $^{232}$U, $^{228}$Th, $^{227}$Ac) have an order of magnitude higher power and energy density than current alpha emitters, and at human spaceflight-relevant doses the $^{227}$Ac and $^{210}$Pb chains need less shield mass than $^{238}$Pu or $^{241}$Am above a few hundred watts. Fusion neutrons could therefore enable nuclear batteries at the kilowatt to megawatt scale and unlock new possibilities for power sources requiring exceptionally high energy density.

Shielding and tritium breeding performance assessment for magnetic mirror fusion devices: a parametric neutronics study with ParaTAN

Hitarth Shah, Tim Bohm, Ed Marriott, Craig Jacobson, Benjamin Lindley

Integrated workflow for DEMO divertor design using Reduced Order Models

D. Valente, C. Poloni, G. Degrassi, R. Ambrosino, F. Subba, D. Marzullo