Cooled Lithium Research Articles

Optimization of the first wall cooling circuits to achieve an efficient coolant flow distribution in the DCLL breeding blanket

In the current design proposal of the Dual Coolant Lithium Lead (DCLL) breeding blanket for the EU DEMO reactor, the first wall is cooled down by a succession of He channels arranged along the poloidal length. In this scheme, since He extracts around 30% of the blanket thermal power, a conservative oversizing of the coolant mass flow rate compelled by highly heterogeneous heat loads, can severely jeopardize the blanket thermal efficiency and the consumption of the auxiliary systems. A meticulous design of the hydraulic network can enhance the cooling system efficiency by ensuring that each channel is fed with a mass flow rate according to its specific needs.In this study, a novel approach employing genetic algorithms has been developed to optimize the geometry of the first wall, encompassing the feeding/collecting manifolds and channels. This optimization aims to replicate the mass flow rate distribution obtained from TOMFLOW (Tool for Optimization of Helium Mass Flow Rate Distribution), a code dedicated to determine the minimum mass flow rate required in each channel to maintain the EUROFER steel structure below 550 °C.

Test blanket modules piping systems calculation method according to RCC-MRx code

The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at equatorial ports located within port plugs (PP). The Pipe Forest (PF) is a network of pipes connecting the Ancillary Equipment Unit (AEU) to the TBMs. Depending of the TBM, the Pipe Forest includes various piping system such as Water Coolant (350°, 188 bars), Helium Coolant (500°, 100 bars) Lithium Lead and Tritium Extraction System. During operating states, the Pipe Forest main function is to withstand loads in normal, incidental and accidental situations according to RCC-MRx code criteria (including design rules for mechanical components for fusion installations). It includes severe thermomechanical loadings due to the thermal expansion added to seismic and vertical displacements events. In addition, the PF layout must satisfy constructability in constraint environments and maximize the space available for man access, which constitutes a real challenge for design.Since no existing RCC-MRx post-processing tool exists off the shelves, a specific methodology for pipe stress analysis is being developed by ITER, CEA and ASSYSTEM. This methodology starts with a post-processing calculation tool of stresses and reaction forces for looping back into the structural calculation. Moreover, a specific extraction module from 3D model to processing tool allows the conversion of CAD data to calculation entry files. Finally, a feedback loop is also re-integrated into the core CAD model through an import module, enabling the tracking of displacements and evaluate the impact of incorporating pipe movements into the analysis.This paper describes the pipe stress analysis methodology and its application for Pipe Forest system design life cycle. It covers the historical evolution of the routing including introduction of expansion loops. More recently, it incorporates the integrated calculation of the entire port cell, spanning from the back of the TBM sets, up to the connection pipes of the vertical shaft.

Progress in design and experimental activities for the development of an advanced breeding blanket

There is no doubt about the interest of achieving as fast as possible the capability to build and operate high performance reactors that finally allow fusion competing in the electricity market. An advanced breeding blanket based on the Dual Coolant Lithium Lead concept with single module segment architecture, designed for the European DEMO, is certainly aligned with such objective. This work describes some recent outcomes of the efforts carried out in the framework of the Prospective R&D Work Package in EUROfusion to develop this line. The evolution of the design has been guided by strategies aimed to achieve an equilibrium between tritium breeding & shielding requirements and mechanical integrity. To enhance the thermodynamic cycle efficiency and reduce the recirculation power by minimizing pressure losses in both the breeding zone and the first wall have been complementary guidelines. Specific models have been created to characterize phenomena like heat transfer and tritium permeation in the breeder channels. The experimental activities, which in general have produced promising results, have consisted in the characterization of different ceramic materials (carbides and oxides) in terms of functional properties at high temperature (as-received and irradiated/implanted samples) and compatibility with PbLi.

Preliminary analysis of in-orbit operation accidents for an ultra-small Lithium-cooled space reactor power system

To study the safety characteristics of an ultra-small (5.2 MWt) lithium-cooled space reactor power system (ULCR) under postulated accidents, a transient analysis code for the ultra-small space reactor coupled with a closed Brayton cycle dynamic conversion and heat pipe radiator was established. A preliminary analysis of four typical postulated accidents occurring at 100 percent power in-orbit operation of the ULCR system was conducted to define postulated problem areas and establish safety requirements. The postulated accidents mainly include (1) reactivity insertions, (2) coolant partial flow loss, (3) turbomachinery failure, and (4) radiator partial failure. The results showed that: a) since the ULCR system was designed to operate at high-temperature conditions (coolant temperature greater than 1500 K), the safety margin of the ULCR system was small. Emergency shutdown control or reduced power operation action should be taken proactively to ensure ULCR's safety in different accidents. b) When the control rod was misoperated and large reactivity was inserted, the maximum temperature of the lithium coolant was very close to the saturation temperature (the difference was 24 K), and the system was operated at extremely high temperatures (the average temperature of the core coolant was 1750 K); c) In coolant flow loss accident, when the coolant flow of the system dropped below 40% of the rated mass flow, the lithium coolant boiling would occur in the primary loop system, and the emergency shutdown control action was required to ensure the safety of the system.

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

Neutronic Assessments towards a Novel First Wall Design for a Stellarator Fusion Reactor with Dual Coolant Lithium Lead Breeding Blanket

The Stellarator Power Plant Studies Prospective R&D Work Package in the Eurofusion Programme was settled to bring the stellarator engineering to maturity, so that stellarators and particularly the HELIAS (HELical-axis Advanced Stellarator) configuration could be a possible alternative to tokamaks. However, its complex geometry makes designing a Breeding Blanket (BB) that fully satisfies the requirements for such a HELIAS configuration, which is a difficult task. Taking advantage of the acquired experience in BB design for DEMO tokamak, CIEMAT is leading the development of a Dual Coolant Lithium Lead (DCLL) BB for a HELIAS configuration. To answer the specific HELIAS challenges, new and advanced solutions have been proposed, such as the use of fully detached First Wall (FW) based on liquid metal Capillary Porous Systems (CPS). The proposed solutions have been studied in a simplified 1D model that can help to estimate the relative variations in Tritium Breeding Ratio (TBR) and displacement per atom (dpa) to verify their effectiveness in simplifying the BB integration and improving the machine availability while keeping the main BB nuclear functions (i.e., tritium breeding, heat extraction and shielding). This preliminary study demonstrates that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% in some of the selected configurations along with a small decrease of 4.9% in TBR. This could even be improved to just a 3.8% TBR reduction by using a graphite reflector. Such an impact on the TBR is considered affordable, and the results presented, although preliminary in essence, have shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution.

Experimental setup for study of mechanical performance of materials in liquid Pb-16Li and first results

Dual Coolant Lithium Lead (DCLL) Breeding Blanket, which is using the eutectic lead-lithium alloy (Pb-16Li) as tritium breeder, neutron moderator and coolant, is being developed in the frame of the EUROfusion project. Flow channel inserts (FCI) are insulating components embedded along the Pb-16Li channels inside the DCLL breeding blanket modules developed to minimize the magnetohydrodynamic pressure drop and heat losses. There is an effort to modify the originally designed steel-ceramic-steel layered structure of FCI by eliminating the steel protective layer leading to simple steel-ceramic structure. In this arrangement, the ceramic is directly exposed to liquid Pb-16Li. Demanding operating conditions of the DCLL blanket, in particular elevated temperatures, may give a rise to thermomechanical stresses leading to component failure. Therefore, the assessment of mechanical behaviour of the material in direct contact with Pb-16Li is of primary interest.To study the mechanical properties of the material exposed to corrosive environment of Pb-16Li at elevated temperatures, a dedicated experimental setup was designed. Three-point flexural test is adopted to evaluate the flexural stress-strain material response. Since the experimental procedure is performed in liquid Pb-16Li at high temperatures with an inert atmosphere, the mechanical test takes place in an isolated testing chamber filled with molten Pb-16Li transferred from the melting chamber prior the mechanical test. A specificity of the presented facility is its capability of material testing of three samples within one testing procedure in Pb-16Li at temperatures up to 650 °C.

Loss of Liquid Lithium Coolant in an Accident in a DONES Test Cell Facility

A Demo-Oriented early NEutron Source (DONES) facility for material irradiation with nuclear is currently being designed. DONES aims to produce neutrons with fusion-relevant spectrum and fluence by means of D–Li stripping reactions occurring between a deuteron beam impacting a stable liquid lithium flowing film implementing the target. Given the hazard constituted by the liquid lithium inventory and the potential risk of reactions with water, air, and concrete eventually resulting in fire events, the Target Test Cell (TTC) is filled with helium and the reinforced concrete walls forming the bio-shield are covered with steel liners. A loss of Li in TTC, due to a large break in the Quench Tank, is postulated, and consequences are deterministically studied. With the TTC liner being water-cooled, the impact of the liner temperature rise following a leakage event is evaluated. Two separate MELCOR code models have been defined for the liquid lithium loop and water-cooled loop and are numerically coupled. The amount of leaked inventory dependent on the implemented safety logic and impact on TTC containment is evaluated. The water pressurization pattern within the liner cooling loop is studied to highlight possible risks of lithium–water/concrete reactions.

The European Dual Coolant Lithium Lead breeding blanket for DEMO: status and perspectives

During the last years CIEMAT has been leading the activities in the European Program to develop an integral breeding blanket with advanced performances to work in a realistic DEMO scenario. This blanket is the dual coolant lithium lead (DCLL) working at a limited temperature in order to allow the use of conventional materials and technologies. The design of this blanket was finished, including the definition of the tritium extraction system and tritium simulations. Then, determined by the selection of other two concepts as driver blankets for DEMO, the focus was put on developing a novel BB still based on the DCLL concept but working at higher temperatures, thus increasing the plant net efficiency. In this work, a summary of the status of the DCLL is presented, together with some ideas for developing an advanced high temperature DCLL in the near future.

Verification and Validation of COMSOL Magnetohydrodynamic Models for Liquid Metal Breeding Blankets Technologies

Liquid metal breeding blankets are extensively studied in nuclear fusion. In the main proposed systems, the Water Cooled Lithium Lead (WCLL) and the Dual Coolant Lithium Lead (DCLL), the liquid metal flows under an intense transverse magnetic field, for which a magnetohydrodynamic (MHD) effect is produced. The result is the alteration of all the flow features and the increase in the pressure drops. Although the latter issue can be evaluated with system models, 3D MHD codes are of extreme importance both in the design phase and for safety analyses. To test the reliability of COMSOL Multiphysics for the development of MHD models, a method for verification and validation of magnetohydrodynamic codes is followed. The benchmark problems solved regard steady state, fully developed flows in rectangular ducts, non-isothermal flows, flow in a spatially varying transverse magnetic field and two different unsteady turbulent problems, quasi-two-dimensional MHD turbulent flow and 3D turbulent MHD flow entering a magnetic obstacle. The computed results show good agreement with the reference solutions for all the addressed problems, suggesting that COMSOL can be used as software to study liquid metal MHD problems under the flow regimes typical of fusion power reactors.

Liquid metal MHD flow influence on heat transfer phenomena in fusion reactor blankets

DEMO reactor is projected as a technological demonstrator and a component test reactor for several key systems such as the breeding blanket. The dual coolant lithium lead (DCLL) is one promising breeding blanket candidate that performs the functions of shielding, part of the cooling and breeding with the same liquid metal fluid. PbLi is also in charge of transporting the bred tritium to the Tritium Extraction System. Structural cooling of the blanket walls is performed by pressurized helium flow which also transfers heat to the balance of plant. Keeping a proper distribution of thermal loads between the two coolants is a key aspect to reach a plant high efficiency. The thermal power transferred from the liquid metal to the helium channels depends on the flow regime. While liquid metal flows under the magnetic field, the magnetohydrodynamic (MHD) phenomena will occur, influencing the flow regime and the velocity profile. The heat is volumetrically deposited in the liquid metal coolant due to the interaction between the highly energetic neutrons originated in the plasma and the coolant. The heat deposition profile is roughly exponential causing temperature differences across the same channel cross section. Temperature gradients induce buoyant effects altering the expected MHD profile and affecting the heat transfer rate from the liquid metal to the surrounding walls cooled with helium. It is of utmost interest to identify the effects of the main flow parameters on the heat transfer coefficient under such complex conditions. The work studies computationally a domain consisting of a liquid region and a surrounding solid region electrically and thermally coupled using latest EU DCLL design characteristics. A fully developed flow is assumed where buoyancy is modelled under the Boussinesq approximation. Joule effect has been considered negligible compared to the neutron deposition. The influence of the buoyant MHD phenomena to the heat transfer coefficient is analysed parametrically with a CFD solver implemented in OpenFOAM®. The parameters investigated here are the Reynolds, Hartmann and Grashof numbers, the Grashof ratio, and the wall conductance ratio (cw).

Study on the performance of hydrogen isotopes permeation sensor in liquid Li–Pb

Liquid Li–Pb alloy is served as neutron multiplier, tritium breeder, and coolant in Dual Coolant Lithium–Lead (DCLL) breeding blankets. In order to monitor the tritium transport continuously and measure the tritium production rate accurately, the concentration of hydrogen isotopes in liquid breeder is an essential factor for tritium measurement. The hydrogen isotopes permeation-based capsule seems to be the most reliable sensor and is simple from the fabrication point of view. To verify the stability and repeatability of hydrogen isotopes concentration measured by permeation capsule in Li–Pb, a system based on a Nb-capsule and a quadrupole mass spectrometry was devised. The sensor would be applied to determine the variation rate of total pressure while a quadrupole mass spectrometry to measure partial pressure values of different hydrogen isotopes. The response time of the sensor is less than 100 s in liquid Li–Pb alloy and the growth rate of pressure has good repeatability in dynamic mode. Besides, calculations based on numerical simulation of the hydrogen isotopes sensor are presented.

Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space.In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation.The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses.

Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre.A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid.

Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development

The present paper describes the experimental campaign executed at the ENEA Brasimone Research Centre aiming at supporting the development of a PbLi/water heat exchanger suitable for the lithium–lead loops of the dual coolant lithium lead and the water cooled lithium lead breeding blankets of the EU DEMO fusion reactor. The experiments were performed in a test section named HERO, installed inside the main vessel of the lead–bismuth eutectic-cooled pool-type facility CIRCE. The test section hosts a steam generator bayonet tube mock-up in relevant scale, which was selected as a promising configuration for DEMO purposes. For the thermal-hydraulic characterization of the component, five tests were executed at different water pressures (6, 8, 12 MPa, two tests at 10 MPa), and liquid metal flow rates (40, 33, 27, 20, 10 kg/s). The experimental outcomes proved the technological feasibility of this novel steam generator and its suitability for the DEMO PbLi loops. The activity was completed with a post-test analysis using two versions of the system code RELAP5. Because the experiments were executed with lead–bismuth eutectic, a scaling analysis is proposed to find the equivalence with PbLi. RELAP5 code was applied to recalculate the experimental data using PbLi as working fluid.

Investigation of the hydrogen production of a laser FUSION driver thorium breeder using various coolants

In this present work, hydrogen production and neutronic calculations of a Laser Inertial Confinement Fusion Fission Energy (LIFE) driven thorium breeder using various coolants have been investigated. In the neutronic calculations for fusion driver power of 500 MWth has been examined with MCNP code. The 95 vol% Flibe or natural lithium 5 vol% TRISO coated ThC fuels have used in the neutronic calculations. Tritium breeding ratio (TBR) has been calculated as 1.08 and 1.19, respectively, for Flibe and natural lithium coolants. The energy multiplication values have been computed as 3.17 and 1.62, respectively, for these coolants. The burnup values with flibe and natural lithium have been obtained as 6 GWd/tM and 22 GWd/tM over 11 and 23 years, respectively. Also, the hydrogen production of a laser fusion driver thorium breeder using steam methane reforming (SMR), high temperature electrolysis (HTE) and sulfur-iodine (S–I) thermochemical water splitting processes have been performed. The highest hydrogen production values with flibe coolant of SMR method have been obtained as ~200 kg/s over 11 years.

Investigation of heat transfer in a steam generator bayonet tube for the development of PbLi technology for EU DEMO fusion reactor

In the frame of the EUROfusion roadmap for the development of the DEMO power plant, a research activity was carried out to develop a Lithium–Lead/water heat exchanger. The component should be capable to remove nuclear heat deposited in the liquid metal of the Dual Coolant Lithium Lead breeding blanket and feeding a steam turbine, ensuring an efficient thermal power conversion to electricity. One of the selected configurations is the steam generator bayonet tube. The HERO test section is an experimental mock-up in a relevant scale of this steam generator, consisting of a bundle of seven double-wall bayonet tubes with a leakage monitoring system. This test section, developed by ENEA at Brasimone Research Center and installed in the main vessel of the CIRCE pool facility, aims to investigate the thermal-hydraulic features of the system, providing a database for thermal-hydraulic system codes validation. An experimental campaign was carried out to demonstrate technological feasibility and performances of the prototypical heat exchanger, suitable as steam generator for the PbLi loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. A post-test analysis has been realized with RELAP5-3D and RELAP5/Mod3.3 codes in order to evaluate code capability in simulating heat transfer in liquid metal side.

Analysis of flow channel insert deformations influence on the liquid metal flow in DCLL blanket channels

Abstract The dual coolant lithium lead (DCLL) is a candidate to be an effective breeding blanket (BB) concept for nuclear fusion technologies. One critical point of this design is the magnetohydrodynamic (MHD) effects involving Lorentz damping force which produces relevant pressure drop in the eutectic flow. In the framework of the European DEMO, the application of sandwich-like steel–alumina–steel flow channel insert (FCI) seems to be the best solution to reduce the pressure drop by electrically decoupling the liquid PbLi from the Eurofer walls. The impact of the FCI on the PbLi velocity profile is analyzed in this work with a CFD solver implemented on OpenFOAM. Under the assumption of non-buoyant fully developed channel flow the temperature map in the channel is computed. Based on the temperature field, the induced deformation is evaluated. The effects of the FCI deformation and possible rupture of the FCI on the velocity profile and on the corresponding pressure drop are then parametrically investigated. Results show that the deformation of the FCI and the possible break in the Hartmann wall do not lead to significant variations in the pressure drop from the case of intact FCI in a wide range of interaction parameters.

Scaling Methodology for Integral Effects Tests in Support of Fluoride Salt–Cooled High-Temperature Reactor Technology

The capability to validate integral transient response models is of critical importance for licensing new reactor designs. The Kairos Power testing program has developed a methodology to design scaled experiments that predict the thermal fluid behavior of the Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR). Such experiments will be used as part of the assessment base of evaluation models supporting KP-FHR safety analysis. The Hierarchical Two-Tiered Scaling (H2TS) methodology was selected for Kairos Power scaling efforts that will be applied to integral effects tests (IETs) for system-level testing. In this paper, the scaling methodology is presented for thermal fluid IETs that will model the KP-FHR primary heat transport system under normal operations and transient conditions involving transition to natural circulation. This paper provides a basis for using surrogate fluids for testing that requires the thermal fluid performance of the KP-FHR primary coolant lithium fluoride–beryllium fluoride [2LiF/BeF2 (Flibe)] to be replicated. Kairos Power intends to use heat transfer oil as a surrogate fluid for Flibe in thermal fluid IETs. This paper demonstrates that this class of surrogate fluids is an acceptable substitute for Flibe salt for some types of scaled IETs and that the principal thermal fluid properties can be properly scaled with minor distortions over the range of conditions expected for both normal and off-normal operating conditions of the KP-FHR.

Status of Pb-16Li technologies for European DEMO fusion reactor

Three of the four breeder blanket concepts currently under investigation for the European DEMO Reactor use the eutectic Pb-16Li as breeder material. Those are the Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL) blankets. Moreover, the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and auxiliary systems is essential. The main functional requirements that Pb-16Li systems have to fulfill are:•to circulate the liquid Pb-16Li through the blanket and ancillaries;•to extract the tritium produced inside the breeder modules from Pb-16Li;•to control Pb-16Li chemistry and to remove accumulated impurities;The present work aims to describe the activities performed in order to achieve the following objectives: i) design and integration of the Pb-16Li loops inside the tokamak building, ii) development and characterization of antipermeation and anticorrosion coatings on structures in contact with Pb-16Li, iii) development and design of an activation products removal system, iv) design of a chemistry control system for Pb-16Li loops, v) performing magnetohydrodynamic analyses taking into account the impact on heat transfer and tritium transport in breeding blankets and performing safety analyses of water/Pb-16Li interaction due to LOCA inside the WCLL blanket.