Fusion Reactor Design Research Articles

Fast low-temperature irradiation creep driven by athermal defect dynamics

The occurrence of high stress concentrations in reactor components is a still intractable phenomenon encountered in fusion reactor design. Here, we observe and quantitatively model a non-linear high-dose radiation mediated microstructure evolution effect that facilitates fast stress relaxation in the most challenging low-temperature limit. In situ observations of a tensioned tungsten wire exposed to a high-energy ion beam show that internal stress of up to 2 GPa relaxes within minutes, with the extent and time-scale of relaxation accurately predicted by a parameter-free multiscale model informed by atomistic simulations. As opposed to conventional notions of radiation creep, the effect arises from the self-organisation of nanoscale crystal defects, athermally coalescing into extended polarized dislocation networks that compensate and alleviate the external stress.

Applicability of alkali beam emission spectroscopy on NSTX-U.

Understanding fast pedestal dynamics and turbulent transport in the edge and scrape-off layer (SOL) plasma of spherical tokamaks is crucial for the design and operation of future fusion reactors. The alkali beam emission spectroscopy diagnostic technique offers a means to measure the absolute electron density radial profile and fluctuation amplitude in these regions. In this study, we demonstrate that injecting a sodium neutral beam radially into the plasma and analyzing the light emission from its 3p-3s atomic transition using near-orthogonal viewing angles allows for accurate measurement of the electron density profile and fluctuations in the National Spherical Torus Experiment (NSTX) Upgrade spherical tokamak. Our findings indicate a peak signal-to-noise ratio of 118 in the pedestal and 12 in the SOL under typical NSTX plasma conditions. The spatial resolution for the electron density profile is estimated to be between 2 and 8mm, while for fluctuation measurements, it ranges from 12 to 15mm.

Optimization of Tokamak Blanket Design to Minimize Neutron Flux on Magnets

Fusion energy, a promising solution to global energy challenges, replicates the same processes that give our sun its energy, notably through deuterium-tritium (D-T) fusion reactions that produce significant energy. This study explores the mechanics and challenges of various fusion reactions, emphasizing the pivotal role of lithium breeder blankets in producing tritium, essential for sustaining D-T fusion in tokamak reactors. The helium-cooled lithium lead (LiPb) blanket design was simulated to optimize tritium breeding and neutron flux management. Results indicated a tritium breeding ratio (TBR) of 1.15, surpassing the self-sufficiency target of 1.1, with further improvements through increased lithium content and blanket thickness. Effective neutron shielding ensured safe operational limits for reactor components. These findings demonstrate the feasibility of achieving self-sustaining fusion reactions, essential for the viability of fusion power as a sustainable energy source. Future research will focus on advanced materials, refined simulations, and enhanced cooling technologies to further optimize fusion reactor designs.

High Current Measurement of Commercial REBCO Tapes in Liquid Helium: Experimental Challenges and Solutions

Recent advances in high-temperature superconductors (HTS) have made them extremely attractive for low-temperature, high-magnetic-field-power applications such as in fusion technology, where the advantages over traditional low-temperature superconductors (LTS) allow for the design of fusion reactors operating in different and more convenient regimes. However, the performance enhancement exhibited by novel conductors poses several challenges for the measurement of their superconducting properties. The high critical currents coupled with the relatively low thermal stability of the conductors and their mechanical fragility render this task a challenge, as the angular anisotropies complicate the experimental setup. In this work, we describe the development of our novel high-current measurement facility, focusing on the solutions introduced regarding critical aspects such as the superconducting leads and the sample holder design. We show how simple but effectively designed solutions can be adopted to combat the complexity of the measurement. The results reported in this work guide the development of a measurement system able to withstand high critical currents (I > 1500 A) at high magnetic fields (µ0H > 12 T) by evaluating the angular response of 4 mm wide short samples (L ~ 7.5 cm) in a robust and reproducible manner.

Development of a boron-containing reduced activation Ferritic-Martensitic (B-RAFM) steel

Steel will be an essential part of any commercial fusion reactor design. Applications in this area involve extreme conditions, imposing particular performance requirements, such as high operational temperature and creep resistance, and also a limitation on the elements that can be used due to the activation that occurs on interaction with irradiation. This work begins with a steel developed for conventional power plant applications, the IBN1 grade developed by IMPACT (a UK consortium of industrial and academic research organisations). This grade has shown excellent properties at high temperature due to high temperature-stable precipitate phases, but contains several elements that would become radiologically active to a degree that is incompatible with the required disposal routes after exposure to the fusion reactor environment. In this study, modifications of the composition are made to remove these elements, and thermodynamic modelling and experimental assessment of the phases that form are undertaken. In this, we have paid particular attention to the prediction of transformation temperatures (to understand if normalisation and tempering can be applied successfully) and the precipitates, to see if suitable phases that are likely to impart creep strength and other desirable properties would be formed. The modifications made include the removal of Nb, Mo, Ni, Co, Cu and Al from the starting alloy, and the substitution of Ta (intended to form carbides, replacing the effect of Nb). Modifications of the amount of retained elemental components, such as C, Mn and Cr, have been made with Thermo-Calc modelling, to ensure preservation of comparable phase transformation temperatures and microstructures. The predicted changes to the alloy are compared to the observations from experimental investigation, finding that tantalum can substitute for niobium in these systems and form similar carbides with similar distribution in the material, and that reduction of Cr to 8 wt% and increase of C to 0.12 wt% raises the Ae4 temperature to allow a high-temperature heat treatment without δ-ferrite formation. While assessment of the mechanical properties of this alloy would be required, the perspectives for these alloys to perform at high temperature that can be inferred from the microstructure are discussed.

Speciation and Diffusivity of Hydrogen in Molten 2LiF-BeF2 for Nuclear Fission and Fusion Applications

The behavior of hydrogen isotopes in molten 2LiF-BeF2 (FLiBe) is of interest to both advanced fission and fusion reactor designs. Tritium (hydrogen-3) is both a fission product and neutron activation product of lithium-6 fluoride in fission reactors which use FLiBe as a heat transfer fluid. In this case, tritium is an undesirable side product. In fusion reactors, FLiBe may be used as a blanket around the plasma which, upon neutron irradiation, generates tritium by the same neutron–Li-6 reaction. Here, tritium is a necessary and desirable product; the tritium is used to continuously fuel the deuterium-tritium fusion reaction. A thorough understanding of tritium’s behavior and residence time is relevant to managing tritium inventories in fission and fusion systems, which is essential for both the safety cases and functional operation of these reactors. Using hydrogen as a non-radioactive surrogate for tritium, we study the speciation and diffusivity of hydrogen in molten FLiBe at various temperatures relevant to reactor operating conditions.Here we present results of electrochemical investigations of hydrogen in FLiBe using cyclic and square wave voltammetry. We introduce hydrogen to FLiBe through 1) LiH additions and as 2) H2 gas, electrochemically desorbed from graphite. The expected solubility of H2 in FLiBe at 1 atm partial pressure is as low as 1 appm. Previous studies saw a higher apparent solubility of H2 at atmospheric pressure (up to 8000 appm). The possible presence of the postulated quasi-stable intermediate BeH2 is explored to explain the high apparent solubility of H2. The two methods of introducing H (as LiH reacting with BeF2 to form BeH2, which decomposes to H2, and the surface reduction of H2 on graphite) are compared to investigate the different species in which hydrogen may exist in molten FLiBe. Additionally, we attempt to quantify the diffusivity of these hydrogen species in FLiBe at 500ºC and 800ºC and compare findings with predicted diffusivities from computational models of Be and H additions to FLiBe. Overall, these studies seek to elucidate the speciation of hydrogen in FLiBe to aid in the management and operation of fission and fusion systems.

GX: a GPU-native gyrokinetic turbulence code for tokamak and stellarator design

GX is a code designed to solve the nonlinear gyrokinetic system for low-frequency turbulence in magnetized plasmas, particularly tokamaks and stellarators. In GX, our primary motivation and target is a fast gyrokinetic solver that can be used for fusion reactor design and optimization along with wide-ranging physics exploration. This has led to several code and algorithm design decisions, specifically chosen to prioritize time to solution. First, we have used a discretization algorithm that is pseudospectral in the entire phase space, including a Laguerre–Hermite pseudospectral formulation of velocity space, which allows for smooth interpolation between coarse gyrofluid-like resolutions and finer conventional gyrokinetic resolutions and efficient evaluation of a model collision operator. Additionally, we have built GX to natively target graphics processors (GPUs), which are among the fastest computational platforms available today. Finally, we have taken advantage of the reactor-relevant limit of small $\rho _*$ by using the radially local flux-tube approach. In this paper we present details about the gyrokinetic system and the numerical algorithms used in GX to solve the system. We then present several numerical benchmarks against established gyrokinetic codes in both tokamak and stellarator magnetic geometries to verify that GX correctly simulates gyrokinetic turbulence in the small $\rho _*$ limit. Moreover, we show that the convergence properties of the Laguerre–Hermite spectral velocity formulation are quite favourable for nonlinear problems of interest. Coupled with GPU acceleration, which we also investigate with scaling studies, this enables GX to be able to produce useful turbulence simulations in minutes on one (or a few) GPUs and higher fidelity results in a few hours using several GPUs. GX is open-source software that is ready for fusion reactor design studies.

Design update of DEMO BOP for HCPB BB concept with an energy storage system

An option for the Intermediate Coupling Design (ICD) of the European Demonstration Fusion Reactor (DEMO) Balance of Plant (BOP) configuration for Helium-Cooled Pebble Bed (HCPB) Breeding Blanket (BB) concept consists of the Primary Heat Transport System (PHTS), the Intermediate Heat Transfer System (IHTS), using HITEC salt as coolant, equipped with a thermal Energy Storage System (ESS), and the Power Conversion System (PCS). The aforementioned BOP configuration is capable of producing electricity continuously during the pulse time (2 h) and dwell time (10 min) of DEMO plant operation.The present study reports on the current (2023) thermal system design of DEMO HCPB ICD BOP for 12 operational states defined by the DEMO Design Central Team (DCT) in 2021. It was shown that the PCS is able to accommodate all of the 12 operational points defined based on the uncertainties of the energy map, provided that an additional bypass line at the outlet of the Divertor Cassettes (DIV-CAS) Heat Exchanger (HX) towards the low-pressure part of the feedwater line (to the condenser in this investigation) is installed in the system. This line is activated only in the cases, where the feedwater flow to the low temperature DEMO heat sources HXs is needed to be higher than that necessary to remove BB power at the Steam Generators (SGs). In all the other cases, the bypass line remains unused. Additionally, perspective improvements of DEMO HCPB BOP regarding thermal ESS, as well as PCS are discussed.

An extensive analysis of SOL properties in high-δ plasmas in ASDEX Upgrade

A set of dedicated H-mode discharges with constant heating power combining Neutral Beam Injection and Electron Cyclotron Resonance Heating have been executed at the ASDEX Upgrade tokamak using a high triangularity magnetic geometry in order to investigate the impact of filamentary transport to divertor and non-divertor components. The evolution of upstream scrape-off layer (SOL) profiles have been correlated with dedicated separatrix quantities, mostly with the turbulence control parameter αt (Eich and Manz 2021 Nucl. Fusion 61 086016) describing the turbulence level at the separatrix. With increasing αt , a broadening of the upstream density profiles in the near-SOL together with the formation of a density shoulder in the far-SOL have been observed. This phenomenon is associated with an enhanced filamentary transport dominating the radial turbulent transport in the far-SOL and confirmed by means of the cooling water calorimetry on non-divertor components. The probe measurements conducted with the ball-pen probe-head mounted on the midplane manipulator and a retarding-field analyzer close to the limiter surface indicate that the key mechanism increasing the radial filamentary transport to the first wall is an increase of the particle flux Γr,fil , caused primarily by the packing fraction fPF,fil and the filament density ne,fil . At the same time, the electron temperature Te and ion temperature Ti measured close to the limiter surface show only small variations above αt > 0.5. Both the filamentary heat flux and the gross erosion derived from the first wall probe measurements reach a magnitude that should be considered in the design of future fusion reactors.

The Irradiation Effects in Ferritic, Ferritic-Martensitic and Austenitic Oxide Dispersion Strengthened Alloys: A Review.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM's corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic-martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

Research on the efficient process-oriented structural optimization method of the large-scale vacuum cryostat for fusion reactors

An efficient optimization design for the large and complex components of fusion reactors is crucial to address the engineering design requirements and further promote technical standardization. Based on research status, current engineering designs for fusion reactors have some deficiencies, such as time and energy wastage, inefficiency, and difficulties in covering the typical ‘multi-variable multi-objective’ design requirements. These are pressing and common problems that urgently need to be overcome. To deal with the aforementioned technical challenges, it is vitally important to design an efficient, precise, and normalized approach that is tailored for the development of future fusion reactors. Therefore, this paper proposes a process-oriented optimization design method, which involves Coupled external parameterized modeling, Experimental points design, Response surface optimization, and Structural integrity validation (CERS), to improve the currently inefficient design methods. And the vacuum cryostat, the largest and complex component of a tokamak, is taken as an example to present the basic procedures of CERS. Firstly, the functions, basic structures, load types, analysis methods, and verification criteria of the cryostat are presented in detail. Then, real-time data interaction between external global parametric variables and ANSYS via coupling is established by CERS, which achieves parametric modeling of the cryostat and efficient experimental point design and optimization analysis with multi-variables and multi-objectives in an automatic way. Subsequently, this study demonstrates the significance and sensitivity of various structural parameters of the cryostat from such objectives as maximum deformation, maximum equivalent stress, and total mass. And the optimal set of its structural parameters is obtained by establishing a mathematical optimization model. Finally, the structural integrity is verified. The results indicate that the optimized cryostat maintains a minimum safety margin of 23% and will not suffer fatigue damage under various load events during its service. Moreover, the nonlinear buckling load multiplier ∅ is 5.4, obtained by analyzing the load-displacement curve of the cryostat according to the zero-curvature criterion. This shows that the designed cryostat is stable enough. The proposed method is simple, efficient, and reliable, and can be applied to both the cryostat and other complex components of fusion reactors in engineering design fields. It has great value of practical technical reference and can further promote the standardization of engineering design technology for future fusion reactors.

Monte Carlo simulations of W-Re alloys under helium ion bombardment

Based on the Monte Carlo simulation method, this study used the SDTrimSP software to explore the sputtering and vacancy distribution of W1−xRex (x = 0, 0.05, 0.10, 0.15, 0.20, 0.25) alloys with (110), (100), and (111) surfaces under helium ion bombardment. Since the surface binding energy in the SDTrimSP software is replaced by cohesive energy, this causes certain difference. So we used LAMMPS to re-scan the surface binding energy of three crystal planes with an index of one at different W-Re alloy ratios to enhance computational accuracy. Subsequently we input this value into the SDTrimSP software to compute the sputtering and vacancy under different incident energies and angles. The results show that (110) crystal plane has the smallest sputtering yield which is related to the surface binding energy. With the increasing concentration of Re atoms in the alloys, the sputtering yield of W-Re alloy linearly increases. Compared with pure W, the difference in sputtering yields and vacancies of W-Re alloys in three crystal planes are not significant. These findings can be valuable when selecting materials for plasma-facing applications in fusion reactor design.

An OpenFOAM solver for multiphysics modeling of fusion reactor design: The nemoFoam code

This paper introduces a multiphysics code, named nemoFoam, designed within the OpenFOAM environment to address the coupling of neutron and photon transport with thermal-hydraulics in nuclear reactor simulations. The code, conceived both for fusion and fission applications, employs a modular approach where the neutronic module is currently based on multi-group neutron diffusion equations, including a mono-kinetic treatment for photons. The thermal-hydraulic module is built on the standard OpenFOAM solver. The paper focuses on presenting the key features of nemoFoam and discussing the structure of the code, the nuclear and thermal-hydraulic models, and the coupling strategy between them.In order to assess the performance of the neutronic module, the code is applied to a fusion case study, indeed a benchmark against the Serpent Monte Carlo code for neutron and photon transport is performed, applying nemoFoam to the Affordable, Robust and Compact (ARC) fusion reactor design. Simulation results demonstrate good agreement with Monte Carlo benchmarks, emphasizing the potential of the code for multiphysics simulations. The modular design of nemoFoam allows users to extend the implemented model equations to study additional phenomena, focus only on selected aspects independently and, potentially, to add new solvers in each module.

ACABLoop simulation tool: Improving the activation prediction of flowing PbLi alloy in support of DEMO fusion reactor design

ACABLoop simulation tool: Improving the activation prediction of flowing PbLi alloy in support of DEMO fusion reactor design

Systematic numerical analysis of ECCD exploring injection conditions with high CD efficiency and current profile controllability in JA DEMO

Systematic Electron Cyclotron Current Drive (ECCD) simulations were performed in JA DEMO plasma (Te=34keV and ne=1.0×1020m−3 at the center) for various injection conditions of EC waves, including wave frequency, injection direction and injection position. Different injection positions surrounding the plasma are comprehensively scanned, including midplane and upper launch on the high/low field side (HFS/LFS) of the plasma. Numerical results indicated that higher harmonic absorptions deteriorate the ECCD efficiency regardless of injection positions. High ECCD efficiencies for each injection position were obtained when avoiding higher harmonic absorptions. The maximum ECCD efficiency of 46kA/MW was obtained for upper LFS launch with an optimal frequency and direction. An important finding from the systematic analysis is that such an efficient ECCD exceeding 40kA/MW is expected for a wide range of injection positions of upper and midplane LFS launch. This result is favorable for fusion reactor design, indicating flexibility in EC port arrangement for efficient ECCD. Current profile controllability in JA DEMO was also discussed in two injection cases: the upper LFS launch and the conventional midplane LFS launch. Driven current density profiles with various injection angles from each injection position successfully cover the regions from the center to mid-radius of the plasma, where external current drive is required to achieve a weak reversed shear. Furthermore, sufficient ECCD efficiencies of 35–46kA/MW at the upper LFS launch and of 35–43kA/MW at the midplane LFS launch are obtained.

A high-density and high-confinement tokamak plasma regime for fusion energy

The tokamak approach, utilizing a toroidal magnetic field configuration to confine a hot plasma, is one of the most promising designs for developing reactors that can exploit nuclear fusion to generate electrical energy1,2. To reach the goal of an economical reactor, most tokamak reactor designs3–10 simultaneously require reaching a plasma line-averaged density above an empirical limit—the so-called Greenwald density11—and attaining an energy confinement quality better than the standard high-confinement mode12,13. However, such an operating regime has never been verified in experiments. In addition, a long-standing challenge in the high-confinement mode has been the compatibility between a high-performance core and avoiding large, transient edge perturbations that can cause very high heat loads on the plasma-facing-components in tokamaks. Here we report the demonstration of stable tokamak plasmas with a line-averaged density approximately 20% above the Greenwald density and an energy confinement quality of approximately 50% better than the standard high-confinement mode, which was realized by taking advantage of the enhanced suppression of turbulent transport granted by high density-gradients in the high-poloidal-beta scenario14,15. Furthermore, our experimental results show an integration of very low edge transient perturbations with the high normalized density and confinement core. The operating regime we report supports some critical requirements in many fusion reactor designs all over the world and opens a potential avenue to an operating point for producing economically attractive fusion energy.

Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors.

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

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 Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

3D neutronic analysis on compact fusion reactors: PHITS-OpenMC cross-comparison

The development of high temperature superconducting tape technology enabled the design of compact fusion reactors, allowing faster development in the field. The reduced size makes even more important the evaluation of neutron flux on all components and at every region, including at the magnet position, to assess the lifetime of the materials and their performances during operations. Compactness, however, introduces new technological challenges, considering the reduced space available for shielding materials and the consequently harsher radiation environment for both structural and functional materials. Many different particle transport codes have been developed for nuclear engineering and fundamental physics purposes, each one relying on different algorithms and nuclear models, with different features and capabilities. In the present work the transport codes PHITS and OpenMC are compared, testing the impact on integral results such as the tritium breeding ratio and on spatially resolved neutron spectra of different nuclear libraries, nuclear models and codes, on a CAD imported 3D geometry of an ARC like fusion machine. The impact of different neutron source shapes is also investigated, from a simple ring-like source to a realistic toroidal plasma distribution, as well as the possibility of performing calculations on a 10°reduced domain, instead of on a full 360°geometry. The uncertainty introduced by each model and geometry choice is analyzed and the computational time required is briefly discussed.

Conceptual design and supporting analysis of a double wall heat exchanger for an ARC-class fusion reactor Primary cooling system

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.