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The water-cooled lead lithium (WCLL) and the dual-cooled lead lithium (DCLL) are two of the breeding blanket (BB) concepts that the EUROfusion consortium is pursuing in the framework of the development of the fusion reactor industrial demonstrator DEMO. Both involve the use of a liquid metal (LM) as working fluid, the lead-lithium eutectic alloy (PbLi), due to its excellent thermal properties and the possibility to serve as both the blanket coolant and tritium breeder and carrier. Unfortunately, due to the high electrical conductivity of LMs, their motion is influenced by the magnetic field used in the reactor to confine the plasma, generating a complex phenomenology which is studied by magnetohydrodynamics (MHD). In this work, a representative prototypical manifold of a BB bottom feeder is investigated for different configurations with the custom phiFoam solver, capable of simulating unsteady, incompressible and isothermal MHD flow. The aim of this study is to investigate which configuration minimizes the flow imbalance in the manifold for the WCLL or in the poloidal breeding zone channels for the DCLL. The distribution of the flow rate among the channels is strongly influenced by the position of the feeding pipe (FP) and by the development of the MHD internal layer near the expansion, which generates important jets close to the lower plate and the upper one, where the channels are attached. The channel aligned with the FP is the one carrying most of the flow, from 55% to 82%, while in the more distant one the flow is almost stagnant, carrying from 17% to 6% of the total flow rate. The total pressure loss is also estimated and its functional dependence on the manifold configuration is discussed. [ABSTRACT FROM AUTHOR]

The breeding blanket is one of the fundamental components of a nuclear fusion reactor and is responsible for the fuel production, generating tritium through neutronic capture reaction between lithium and neutrons. Lithium is a liquid PbLi alloy and the helium formed as reaction by-product can coalesce into bubbles, generating a two-phase mixture with a high-density ratio ( η ρ ∼ O 5 ). These bubbles can accumulate and stagnate within the blanket channels with potentially harmful consequences. In this work, the interIsoFoam solver of OpenFOAM v2012 is used to simulate bubble motion for a two-phase mixture representative of the He-PbLi system to test its potential for future developments in the field of fusion. In a first phase, several traditional benchmarks were carried out, both 2D and 3D, and considering the two variants of the VOF method implemented in the solver, isoAdvector and plicRDF. Subsequently, He bubbles of different diameters rising in liquid PbLi ( η ρ = 1.2 × 10 5 ) were analysed to investigate different regimes. For a Eötvös number (Eo) greater than 10, it was possible to recreate the axisymmetric, skirted, oscillatory regimes and the peripheral and central breakup regimes. For Eo < 10, non-physical deformations of the interface are observed, probably generated by spurious velocities that have a greater impact on the solution for very small bubbles and rising velocities. [ABSTRACT FROM AUTHOR]

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from H a = 500 − 3000 , confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range. [ABSTRACT FROM AUTHOR]

The transport, permeation and retention of tritium inside nuclear fusion reactors is a topic of great interest due to the scarcity of the isotope, its ease of diffusion through materials and its radioactivity. An accurate balance of tritium is needed in all the fuel cycle, and each loss term needs to be evaluated in detail. In this context, reliable and flexible tools to evaluate tritium permeation and retention are a necessity for the development of fusion technologies. This work presents the OpenFOAM PermeAtion Solver for Tritium Analysis Foam (pastaFoam) solver and two custom boundary conditions, along with a series of verification and validation cases. The solver inherits all capabilities of the base solver, chtMultiRegionFoam, and is capable to simulate hydrogen transport in coupled fluid–solid systems in the presence of hydrogen traps, under diffusion limited regime or accounting for surface effects. All features are tested against analytical solutions and results are compared with other tritium transport codes. Agreement with experimental data is aligned with results of numerical benchmarks from literature. • Development of an OpenFOAM multi-region solver for hydrogen and tritium transport. • Development of boundary conditions for diffusion limited and general regime. • Promising verification & validation results against analytical & experimental data. • Solver for migration, permeation and retention analysis in fusion reactors. [ABSTRACT FROM AUTHOR]

Several systems in nuclear fusion reactors utilize liquid metals as working fluids and the design of these systems cannot overlook the magnetohydrodynamics effects arising from the interaction between the electrically conductive fluid and the magnetic fields used to confine the plasma, since these effects significantly influence the flow features. In this context, rigorous studies and research activities are imperative to provide high–quality numerical data and develop precise predictive numerical tools. This work introduces two OpenFOAM magnetohydrodynamics solvers and outlines their respective validation processes. The mMRF solver can simulate single–phase, incompressible MHD flow for multiple electro–coupled domains. Meanwhile, the mIF solver is capable of simulating two–phase MHD flow involving incompressible and immiscible fluids. The mMRF solver has demonstrated outstanding results in simulating classical 2D benchmarks up to high magnetic field intensities, while the mIF solver proved its ability by estimating, with a discrepancy of less than 10 %, the velocity of a bubble rising within a liquid metal under an imposed magnetic field. [ABSTRACT FROM AUTHOR]

• Development of two well-defined cores with the increase of Hartmann number. • A correction factor between pressure gradient of annular and an equivalent channel is defined. • Approximate estimate of manifold pressure drop through the correction factor. The Breeding Blanket is a fundamental component of a nuclear fusion reactor and the Water-Cooled Lead Lithium (WCLL) blanket is one of the possible solutions proposed. In this concept, liquid lithium-lead eutectic alloy (PbLi) serves as tritium breeder, tritium carrier and neutron multiplier. The liquid metal is distributed within the breeding zone by two co-axial rectangular channels and, interacting with the reactor magnetic field, leads to the arising of MagnetoHydroDynamic (MHD) effects. In this work, the general-purpose CFD code Ansys CFX 18.2 is used to study this uncommon configuration, modelled as a prototypical square annular channel. The study covers a wide range of magnetic field intensity (up to H a = 2000) and two values for the wall conductance ratio (c w = 0 and c w = 0.1) representing, respectively, the ideal insulated case and one more closely approaching the WCLL manifold actual conditions. For both these scenarios, characteristic flow features and their evolution with increasing magnetic field are discussed. A correlation is found linking the pressure loss in the studied configuration and an equivalent square channel through a corrective factor ε , which exhibits an asymptotic behavior for H a > 1000 equal to ε c w = 0 ≅ 2.44 and ε c w = 0.1 ≅ 1.12. [ABSTRACT FROM AUTHOR]

In the fight against climate change through the progress of fusion technology, the Affordable, Robust and Compact (ARC) fusion reactor design is under development at Commonwealth Fusion Systems (CFS) and Massachusetts Institute of Technology (MIT)- Plasma Science and Fusion Center, in collaboration with the Ente Nazionale Idrocarburi (ENI) S.p.A. The reactor features demountable superconducting toroidal field coils and a replaceable vacuum vessel immersed in a Fluorine Lithium Beryllium (FLiBe) molten salt blanket. The low-pressure molten salt cools the divertors and the blanket, shields the magnets and acts as tritium breeder and tritium carrier. Tritium must be recovered efficiently from the salt to fuel the reactor, minimizing its inventory and its migration outside the FLiBe loop. The current work is aimed at presenting a preliminary design of a Double-Wall Heat eXchanger (DWHX) to be installed in an ARC-class reactor primary cooling system. The main feature is the presence of a flowing sweep gas gap which captures tritium and prevents its diffusion toward the secondary system. For the DWHX, thermal-hydraulic and tritium transport analyses were performed, considering a supercritical water Rankine power conversion system on the secondary side to carry out the calculations. To assess tritium extraction capabilities of the DWHX, the tritium extraction efficiency has been evaluated. Calculation results highlighted how the current configuration is not adequate to remove tritium, in fact the tritium extraction efficiency is very low. On the other side, it seems to act adequately as tritium permeation barrier. [ABSTRACT FROM AUTHOR]