Breeding Blanket Concepts Research Articles

Thermomechanical and Thermofluid-Dynamic Coupled Analysis of the Top Cap Region of the Water-Cooled Lithium Lead Breeding Blanket for the EU DEMO Fusion Reactor

In the EU, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept is one of the candidates for the design of the DEMO reactor. From the past campaign of analysis emerged that the thermal-induced stress led to the failure in the verification of the RCC-MRx structural criteria. Hence, in this paper the classic conceptual design approach, based on a pure FEM thermal and structural analysis, is compared to a coupled thermofluid-dynamic/structural one. Even though the coupled approach requires tremendous modelling effort and computational burden, it surely allows determining the thermal field with a higher level of detail than the FEM analysis. Therefore, in this work, the focus is put on the impact of a more detailed thermal field on the DEMO WCLL BB global structural performances, focusing on the Top Cap region of its Central Outboard Blanket segment. The obtained results have allowed confirming the soundness of the design solution of the Top Cap region, except for concerns arising on the mass flow rate distribution. Moreover, results have shown that, globally, the pure FEM approach allows for obtaining more conservative results than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible adopting the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalise the most promising.

Exploratory study of the EU-DEMO Water-Cooled Lithium Lead breeding blanket behaviour in case of loss of cooling capability

Within the framework of the European Roadmap to the realization of fusion energy, a strong international cooperation is ongoing to develop a Breeding Blanket (BB) system for the EU-DEMO reactor. Although it is still to be decided whether the DEMO in-vessel components should perform any safety function, the pursuing of robust blanket concepts able to handle upset and accidental loading conditions has been always seen as good practice in fusion reactor engineering to enhance the inherent plant safety performances. Amongst the several classes of events that might challenge the BB structural integrity, the large Loss of Coolant Accident is one of the most relevant because it usually leads to a fast loss of cooling capability of the structures. Due to the characteristic of the tokamak assembly, the behaviour of each blanket segment during a sudden loss of cooling capability does not depend only upon distinguishing features of the component itself. In fact, the overall transient can be governed by conditions established in surrounding elements, like adjacent blanket segments and vacuum vessel, as well as by the plasma shutdown strategies adopted to protect the reactor.The scope of the activity herein presented is to make a preliminary assessment of the intrinsic capability of EU-DEMO tokamak architecture to cope with the loss of cooling in the Water-Cooled Lithium Lead (WCLL) BB concept. Evaluation of BB thermal field in short and medium term under simplified, yet conservative, assumptions was carried out for four transient scenarios with the aim of investigating the response of the structure in case of: a) fast or soft plasma shutdown, and b) different blanket cooling schemes. Moreover, the WCLL BB thermo-mechanical response in the most critical time steps has been assessed. The obtained results shall help for future decisions on safety systems/action to be implemented to cope with accidents.

Theoretical evaluation of the tritium extraction from liquid metal flows through a free surface and through a permeable membrane

Effective tritium extraction from PbLi flows is a requirement for the functioning of any PbLi based breeding blanket concept. For a continuous plant operation, the removal of the tritium dissolved in the PbLi has to be performed in line and sufficiently fast. Otherwise, tritium inventories in the liquid metal, start-up inventories and buffer inventories would be excessive from the safety point of view. Moreover, a slow response of the tritium extraction systems could also compromise the tritium self-sufficiency of the plant. A promising solution to this problem is to use highly permeable membranes in contact with the PbLi flow to promote the extraction via permeation. This technique is usually known as Permeation Against Vacuum (PAV). As an alternative, tritium could be extracted directly by permeation through a fluid free surface (FS) in contact with vacuum. In both configurations, the dynamics of tritium transport is ruled by a combination of convection, diffusion and surface recombination. In this paper, the tritium extraction processes in the FS and PAV configurations are studied in detail. For the first time, general analytical expressions for the extraction efficiency are derived for both techniques in a Cartesian geometry. These expressions are general in the sense that they do not impose any kind of assumption concerning the permeation regime of the membrane or the fluid boundary layer. The derived expressions have been used to analyze numerically the response of both configurations in a close loop system, such as the one of DEMO. The presented methodology allows comparing the FS and PAV configurations, assessing in which conditions one will be behave better than other.

Neutronics assessment for Detailed Helium-Cooled And Water-Cooled DEMO Breeder Blanket Concepts

Neutronics assessment for Detailed Helium-Cooled And Water-Cooled DEMO Breeder Blanket Concepts

A Transient 3-D CFD Model for the Simulation of Forced or Natural Convection of the EU DEMO In-Vessel Components

As the EU DEMO reactor will act as a Component Test Facility for the breeding blanket (BB), it is foreseen that the different BB concepts will be tested throughout the plant’s lifetime. The maintenance of all the in-vessel components (IVCs), as for all D-T fusion machines, must be carried out employing remote handling (RH) technology, as the structural materials will be activated by the neutrons. The maintained segment and possibly other nearby segments cannot be actively cooled and will heat up due to the decay heat. For these reasons, alternative cooling strategies need thus to be investigated to ensure that the BB segment will cool down within the limits required by the RH in a reasonable amount of time. In the present work, two possible cooling options are investigated for the case of the Water-Cooled Lithium-Lead BB concept. One is based on the passive cool-down by natural convection of the BB segments, whereas the second one relies on a forcing flow of cool air on the BB surfaces. A computational fluid dynamics (CFD) approach has been used to study the different options for performing transient analyses through the Star-CCM+ commercial code.

Entropy generation minimization-based heat transfer and mass transfer study of rectangular packed-bed filled with spherical particles

The study focuses on tritium-extraction and heat-extraction characteristics of the lead lithium ceramic breeder (LLCB) blanket module, an Indian tritium breeding blanket concept design in the International Thermonuclear Experimental Reactor (ITER). The present study focuses on the mass transfer phenomenon and entropy generation minimization (EGM)-based heat transfer analysis of the packed bed of a fusion reactor considering tritium isotopes in helium gas. For temperature-dependent convection-diffusion of tritium in helium gas and its permeation through the solid structure, a multi-region model in OpenFOAM using the locally varying porous medium approach is developed for studying the coupled multi-physics phenomena. For heat extraction behavior, a parametric study-based analysis is done using the EGM approach that involves irreversibility due to heat transfer, mass transfer, viscous, porosity effect, and heat generation rates. It is found that a lower Reynolds number (Re), and higher non-dimensional heat generation number (β), and non-dimensional species generation number (ϛ) result in a higher tritium extraction. In contrast, due to minimum total entropy generation, higher Re, and lower β and ϛ are preferred. Random packing arrangement as compared to BCC and FCC arrangements gives better tritium extraction and uniformity. However, the FCC arrangement is found as a better choice based on entropy minimization. Diagonally-opposite as compared to same-side inlet-outlet configuration extracts tritium more effectively and also found to be a better configuration from the EGM approach. The present work is significant since, for the breeder blanket, it presents the implementation of Sievert law boundary conditions and the EGM model for the first time in the literature.

Research on hot isostatic pressing process of CFETR WCCB components

Chinese Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak reactor of equivalent size and complementary functionality to that of ITER. The water-cooled ceramic breeder (WCCB) breeder blanket designed by the Institute of Plasma Physics, Chinese Academy of Sciences is one of the candidate breeder blanket concepts for CFETR. As an important part of the fusion reactor, the WCCB directly determines the safe operation of the fusion reactor. In the WCCB manufacturing technology, the hot isostatic pressing (HIP) process is applied to the manufacture of WCCB due to its unique technical characteristics. This paper introduces the process research and inspection during the development of WCCB components, and the results will be the guidance for the manufacture of WCCB Components.

Development and application of an alternative modelling approach for the thermo-mechanical analysis of a DEMO water-cooled lithium lead breeding blanket segment

In the frame of the EUROfusion research activities devoted to the design of the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead BB (WCLL) concept is one of the candidates currently assessed in EU. To this end, an intense research campaign is ongoing to develop a robust geometric configuration for the WCLL BB Central Outboard Segment (COB). Since the current reference design of the WCLL COB segment is not mature enough to allow a full thermal-hydraulic assessment, an alternative procedure aimed at obtaining a thermal field for the whole segment without performing its complete thermal-hydraulic analysis is presented and applied in this work. The scope of the work is to obtain a thermal field for the whole WCLL COB segment detailed enough to ensure a reliable prediction of the secondary stress spatial distribution, which is pivotal for the segment structural analysis.The procedure is articulated in three steps. Firstly, a thermal analysis of the equatorial region of the WCLL COB is performed, in order to calculate the thermal field in the most loaded region of the segment. Then, a multi-region interpolation is carried out in order to obtain a set of polynomial functions of the radial and toroidal variable, accurately representing the thermal field of the assessed equatorial BB slice. Finally, the calculated functions are applied to the whole COB to impose a realistic thermal field to the segment for the prediction of the stress, strain and displacement fields.Since the mesh for the analysis of the whole segment is expected to be coarser than that used for the assessment of its equatorial region, in this paper a mesh independence assessment of the interpolating functions is reported. In particular, their predictive power both in terms of temperature and secondary stress (i. e. thermal-induced stress) is assessed taking into account a decreasing level of detail of the adopted spatial discretization grid.The obtained results, widely reported and discussed hereafter together with the models and the adopted assumptions, show a great predictive power of the adopted polynomial functions, with a good mesh independence of the calculated temperature and stress values.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

Analysis of EU-DEMO WCLL Power Conversion System in Two Relevant Balance of Plant Configurations: Direct Coupling with Auxiliary Boiler and Indirect Coupling

Among the Key Design Integration Issues (KDIIs) recently selected for the DEMOnstration Fusion Power Plant (DEMO), the operation of the Balance of Plant (BoP) Power Conversion System (PCS) has been recognized as a crucial aspect, due to the typical pulsed regime characterizing the fusion power plant. In the framework of the DEMO Water-Cooled Lead-Lithium Breeding Blanket (WCLL BB) concept, three BoP solutions have been recognized to be able to overcome this issue. They rely on different coupling options between the Primary Heat Transfer Systems (PHTSs) and the PCS: an Indirect Coupling Design (ICD) with Intermediate Heat Transport System (IHTS) and Energy Storage System (ESS), a Direct Coupling Design (DCD) with AUXiliary Boiler (AUXB), and a DCD with small ESS. The present paper deals with a preliminary feasibility assessment of the first two solutions. The analysis, carried out with the GateCycleTM code, referred to a preliminary design phase, devoted to the sizing of the main components, and to a second phase focused on the cycle optimization. The study demonstrated the feasibility of the two BoP concepts. They are able to produce a satisfactory average electric power (>700 MW) with an acceptable average net electric efficiency (33.6% for both concepts). For each solution, the main strengths and weaknesses are compared and discussed.

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the water-cooled lithium–lead breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone Research Centre, Italy. The facility is able to qualify gas–liquid contactor (GLC), permeator against vacuum and liquid-vacuum contactor technologies at different temperatures, lithium–lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.

Evaluation of the thermal-hydraulic performances of a once-through steam generator in nuclear fusion applications

After decades of operation in nuclear power plants, Once-Through Steam Generators (OTSGs) were recently proposed for nuclear fusion applications. In particular, they are supposed to be installed in the primary cooling systems of the European Union Demonstration fusion power plant (EU-DEMO). One of the key reactor components is the Breeding blanket (BB). Among the BB concepts that are currently under study, Water-Cooled Lithium-Lead (WCLL) option was considered for this work. The WCLL blanket is divided in two main subsystems, the breeder zone (BZ) and the first wall (FW), each one provided with an independent cooling circuit, named Primary Heat Transfer System (PHTS). Thermal power removed from BB by BZ and FW PHTS is driven to the Power Conversion System (PCS) to be converted into electricity. The thermal coupling is ensured by two OTSGs per system. At the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome, a simulation activity was carried out to understand the component thermal-hydraulic behavior during DEMO normal operations. For calculation purposes, a full model of the steam generator was prepared by using a modified version of RELAP5/MOD3.3 system code. The computational activity performed allows to preliminary characterize the OTSG thermal-hydraulic performances during both pulse and dwell phases.

Radiological characterization of ceramic materials considered for the HT-DCLL DEMO reactor

The recent evolution of the Breeding Blanket (BB) Dual Coolant Lithium-Lead (DCLL) concept for the European DEMO from a low temperature (LT) multi-module segment (MMS) approach, limited by the thermal range tolerated by EUROFER, to an advanced high temperature (HT) single-module segment (SMS) architecture, is allowed by the consideration of ceramics and composite materials as main structural material rather than just as thin Flow Channel Inserts (FCI). The amount of the ceramic as main structure instead than as a FCI in the whole reactor could be of several tonnes generating concern regarding the radiological behaviour of such component and the consequent generated waste. For that, a preparatory assessment considering ideal “pure” ceramics has been performed towards a preliminary selection of the structural material, among others, under the criteria of maintenance operations and waste management. To complement such study additional analyses have been carried out considering not only the intendent element but also dopant and impurities which often give rise to significant additional activation. For both the theoretical compositions as well as the industrial ones, with a certain amount of impurities, activation calculations have been performed using the ACAB inventory code. Hence, total beta-gamma activity, specific activity for different nuclides, decay heat and surface gamma dose rate have been analysed with reference to the IAEA and SEAFP-2 standards for waste and handling classifications and to the specific regulations of the near-surface repository El Cabril (Córdoba, Spain). According to El Cabril regulation, pure SiC and TiC would be the best of the options considered since they would be accepted in Level 1 LILW (under detritiation for SiC). This may not be true for industrial compositions with impurities (Level 2), nor according to other standards. Pure zirconia is also a promising option, so further work is ongoing for zirconia and doped zirconia materials.

Development of a Compact Multivariable Sensor Probe for Two-Phase Detection in High Temperature PbLi–argon Vertical Columns

Liquid gas two phase flow is a common occurrence in various industrial applications. For nuclear fusion applications with a Li based breeder, existence of two phase flow may lead to critical issues including decreased tritium breeding ratio, generation of hot spots and improper nuclear shielding. Additionally, a very large density ratio of liquid metal to gas mandates relevant experiments towards development and validation of software tools. PbLi has gained immense focus for its various advantages and is utilized in several breeding blanket concepts. In view of above mentioned requirements, a two phase detection tool is imperative for liquid PbLi environment. For liquid metal applications, electrical conductivity based probes are most suitable in terms of ruggedness, fabrication ease and operational simplicity. However, corrosive nature and high operational temperature for PbLi put severe demands on electrical insulation, a foremost requirement for electrical conductivity based schemes. In this study, a multivariable probe based on electrical conductivity and temperature measurement schemes is developed using Al2O3 coating as electrical insulation. Developed probe is validated in PbLi/Ar vertical column with bulk PbLi temperature upto 400 degC and time averaged void fraction upto 0.95 covering flow regimes from dispersed bubbly flow upto in box loss of coolant accident. Developed probe provides high reliability and excellent temporal resolution towards individual bubble detection through electrical conductivity based principle alongwith simultaneous two phase temperature trends. Present paper provides details about sensor probe fabrication, calibration, two phase test facility, void fraction estimations, bubble frequency and bubble residence time estimations alongwith critical observations from preliminary experimental investigations.

Structural Assessment of the European DEMO Water-Cooled Lithium Lead Breeding Blanket Central Outboard Segment Under Remote Maintenance Loading Conditions

Within the framework of the European research activities devoted to design the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead (WCLL) BB concept is one of the candidates currently assessed in EU. One of the main issues connected to its design is represented by the remote maintenance (RM) operations. In particular, previous thermo-mechanical analysis highlighted a Ductile to Brittle Transition Temperature (DBTT) shift above the room temperature within the WCLL central outboard (COB) segment during normal operation. This may pose problems during the RM operations foreseen in the BB replacement phase. Indeed, within those WCLL COB segment’s regions where the DBTT shifted above the room temperature, the stress level arising during RM operations may be able to promote crack growth and migration resulting in a fragile fracture event. Then, in this work, the structural assessment of the DEMO WCLL COB segment under RM scenarios is performed, in order to provide the fracture mechanics analysis with the necessary input in terms of stress spatial distribution and to understand if the deformation occurring during RM can be accommodated by the available space reservation. Several potential RM scenarios have been considered, taking into account also the potential occurrence of a seismic event or the necessity to remove the WCLL COB segment with the liquid breeder not drained. The study has been performed adopting a numerical approach based on the Finite Element Method and adopting the quoted Abaqus code. The obtained results are quite encouraging and will allow, in the follow up, the refinement of the RM plan foreseen for the WCLL BB.

Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

Multiphysics Optimization for First Wall Design Enhancement in Water-Cooled Breeding Blankets

The commercial feasibility of the first fusion power plant generation adopting D-T plasma is strongly dependent upon the self-sustainability in terms of tritium fuelling. Within such a kind of reactor, the component selected to house the tritium breeding reactions is the breeding blanket, which is further assigned to heat power removal and radiation shielding functions. As a consequence of both its role and position, the breeding blanket is heavily exposed to both surface and volumetric heat loads and, hence, its design requires a typical multiphysics approach, from the neutronics to the thermo-mechanics. During last years, a great deal of effort has been put in the optimization of the breeding blanket design, with the aim of maximizing the tritium breeding and heat removal performances without undermining its structural integrity. In this paper, a derivative-free optimization method named “Complex method” is applied for the design optimization of the European DEMO Water-Cooled Lithium Lead breeding blanket concept. To this purpose, a potential performances-based objective function, focusing on the maximization of the tritium breeding, is defined and a multiphysics numerical model of the blanket is developed in order to solve the coupled thermo-mechanical problem, while the optimization algorithm leads the design towards a minimum optimum point compliant with the prescribed requirements. Once the optimized design is obtained, its nuclear and thermo-structural performances are assessed by means of specific neutron transport and multiphysics simulations, respectively. Finally, the structural integrity is verified by means of the application of the RCC-MRx design criteria. • A tool for First Wall optimization in water-cooled breeding blankets is developed. • The optimization is performed using the gradient-free method named “Complex method”. • An objective function aiming at tritium breeding maximization is defined. • A FE model is used to check the compliance with the thermo-structural requirements. • The optimized design is finally assessed by proper, more detailed numerical analyses.

Overview of Thermal Hydraulic Optimization and Verification for the EU-DEMO HCPB BOP ICD Variant

When progressing from the International Thermonuclear Experimental Reactor (ITER) to the Demonstration Fusion Reactor (DEMO), a system for transferring plasma heat exhaust to a power conversion system is necessary for the so-called Balance of Plant (BOP). During the preconceptual phase of the EU-DEMO project, different BOP concepts were investigated in order to identify the main requirements and feasible architectures to achieve that goal in the most efficient way. This paper comprises the investigations performed during the DEMO preconceptual design phase (p-CDP) and compares the different variants. The main aspect was focused on the helium-cooled pebble bed (HCPB) breeding blanket (BB) concept. After all assessments were performed, the indirect coupled design (ICD) was chosen as the reference configuration for the DEMO HCPB BOP for further development and optimization. The ICD provides decoupling using a molten salt storage loop, which accumulates thermal power during plasma pulses that are released during dwell periods. The work is supported by simulations using design codes EBSILON and MATLAB/SIMULINK, providing the basis for the next design phase.

Heterogeneous design and mechanical analysis of HELIAS 5‐B helium‐cooled pebble bed breeding blanket concept

One of the most challenging objectives of the European research concerning nuclear fusion technology, promoted by the EUROfusion consortium, is to bring stellarator-type nuclear fusion devices to maturity. To this purpose, studies on a large HELIcal-axis advanced stellarator (HELIAS), extrapolated from Wendelstein 7-X and based on a 5-fold symmetry (HELIAS 5-B), are currently ongoing. The HELIAS 5-B stellarator reactor will be endowed with a breeding blanket (BB) system to allow for the self-sustainability of the nuclear fusion reaction and make it suitable for electricity generation. In this paper, we present the first ever heterogeneous mechanical design and the preliminary structural assessment of a bean-shaped ring of a HELIAS 5-B BB sector. The proposed mechanical design, which is based on the helium-cooled pebble bed (HCPB) BB concept and developed according to the “sandwich” architecture, foresees an actively cooled segment box connected to a back-supporting structure equipped with manifolds. The internal region (breeding zone) is reinforced by actively cooled steel plates. The proposed heterogeneous design was checked against nominal loads and an in-box loss of coolant accidental scenario, which is a typical design driver for BBs. The assessment has been performed according to the RCC-MRx structural design code. Our results are herewith presented and critically discussed, focusing on the potential follow-up of the HELIAS 5-B HCPB BB design.

The influence of MHD boundary layers on tritium permeation in PbLi flows for fusion breeding blankets

In PbLi based breeding blanket concepts, tritium is produced inside the liquid metal and drag out of the reactor by the liquid metal flow. However, undesired permeation through the channels and pipes walls occurs spontaneously since tritium naturally diffuses in the opposite direction of the concentration gradient. This way tritium can reach the blanket coolant circuit or even the exterior with an impact on the tritium self-sustainability and the safety of the plant. Similarly to heat transfer processes, permeation through the walls in the interface between the flow and the steel is mostly affected by the dynamics of the boundary layers. This is ruled by the electrical coupling between the moving conductor and the conducting walls as a result of the Magnetohydrodynamics (MHD) interactions which dominate the flow dynamics. In this work, the connection between the MHD forces and tritium transport is numerically studied using the simulation platform ANSYS-Fluent. The velocity profiles of a PbLi test channel have been firstly computed in a wide range of Hartmann numbers from 102 to 104. These velocity profiles are then applied to a 3D tritium transport model developed with the customization capabilities of the same platform. A series of tritium transport simulations are carried out considering different permeation regimes: surface-limited, diffusion-limited and intermediate regimes. The development of the concentration boundary layers along the channel is studied in different permeation regimes, magnetic fields and velocity fields. This has allowed correlating the Sherwood number (Sh) with the Hartmann (Ha), Reynolds (Re) and permeation numbers (W).