Cooled Lithium Lead Research Articles

Physics and technology maturity level required for the K-DEMO design points

The conceptual design study for the Korean fusion demonstration reactor (K-DEMO) has been conducted, placing an emphasis on the imperative to reduce the reactor's dimension as a means to cost minimization. The design space of K-DEMO was delineated based on maturity level of physics and technology. Advanced technology features such as a use of a tungsten carbide (WC) shield, a tritium breeding blanket concept of He cooled lithium lead (HCLL), the maximum allowable magnetic field at the TF coil, Bmax = 16 T with a Nb3Sn superconducting material, etc. were adopted. With specified design criteria, including a net electric power ≥ 300 MW, a fusion gain, Q > 20.0, a neutron wall loading < 2.0 MW/m2, an indicator of divertor power handling capability (ratio of power to divertor to major radius), Pdiv/R0 < 25 MW/m, and the capability for steady-state operation, a design space of the K-DEMO was established based on the energy confinement scaling law of IPB98[y,2] under physics level of Greenwald density fraction, ne/nG < 1.2, normalized plasma beta (ratio of plasma pressure to magnetic pressure normalized by plasma current divided by the product of minor radius and toroidal magnetic field), βN < 3.0, confinement enhancement factor, H < 1.3, and a direct cost ≤ 7.5 B$. After an exploration of system parameters, prospective design points for K-DEMO were identified, characterized by a major radius, R0 ∼ 6.8 m, an aspect ratio, A = 3.1, a toroidal magnetic field at plasma center, BT ≥ 6.5 T, and a fusion power, Pfusion ∼ 1,500 MW. When β-independent energy confinement scaling law was applied, the design points were accessible with smaller ne/nG, βN, H, Pfusion, and larger BT. From a sensitivity analysis of the minimum major radius to the input parameters, strong sensitivities to Greenwald density fraction, ne/nG, normalized plasma beta, βN, confinement enhancement factor, H, edge safety factor, qedge, and elongation, κ were found. Additionally, the operational envelope in physics and technology parameters was established with system parameters associated with the design points.

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.

Development of the high temperature PbLi experimental facility for CFETR

Chinese Fusion Engineering and Test Reactor (CFETR) aims to demonstrate the fusion energy production up to 200MW, and finally reach commercial electricity power level 1GW. As one key component, the blanket is in charge of tritium breeding, neutron shielding and energy conversion. The liquid breeder blanket has many important advantages of simpler structure, tritium extraction and fuel replenishment online, as well as the high thermoelectric conversion efficiency. Moreover, the supercritical carbon dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket has been proposed for CFETR. However, the key issue of MagnetoHydroDynamics (MHD) needs further investigation, especially for the experiments. Therefore, the high temperature PbLi experimental loop is under development, which mainly includes three systems of PbLi, oil and water. The main components include the mechanical pump, main heater, cold trap as well as the superconducting magnet. At the test section, the mass flow rate, temperature and magnetic intensity will be up to 32 kg/s, 700 °C and 3T, respectively. In addition, the ultrasonic doppler velocimeter and potential probe instruments will be adopted to obtain the flow field. The main goal is to establish flow and heat transfer models, which are the critical input for the liquid breeder blanket design and optimization.

Thermal hydraulic assessment on the full banana model of COOL blanket for CFETR

The Supercritical Carbon Dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket is under development for Chinese Fusion Engineering and Test Reactor. The thermal hydraulic assessment plays an important role for the comprehensive performance evaluation on the fusion blanket among the multi-physics fields. As the fusion reactor will enter into the engineering construction stage, it is important to study the thermal hydraulics performance on basis of the full model. Because it can accurately check the heat removal capability and thermoelectricity conversion efficiency, as well as provide essential input for the other physical fields. In this demand-driven, the analyses and optimization on the cooling system are put into priority on basis of the full banana model, including the manifold design and inlet/outlet pipes locations. Finally, the coolant pressure drop is highly reduced and the mass flow distribution becomes much more uniformly. For the S-CO2, 82.3% of the total mass flow rate is distributed into the key component first wall, and this is beneficial to face the high radiation heat flux. Besides, under different level of heat flux, the required total mass flow rate and pressure drop are obtained on premise that the coolant has enough ability to safely remove the heat away. For the Lead–Lithium (PbLi), the distribution of mass flow rate is designed as ‘ladder’ shape to adapt the unevenly spatial distributed nuclear heat along the radial direction, and the ratio is 8:2:1. Furthermore, the first law of thermodynamics is adopted for the trade-off analysis, which converts the total pressure drop of the two coolants into the pumping power, and it occupies only 1.3% of the total thermal power. This provides accurate and valuable data for the primary and secondary loop design, as well as the economic assessment on the fusion reactor. Finally, the Two Dimensional thermal hydraulic model containing the detailed layouts of different materials is used to study the coupling heat transfer effects between PbLi and S-CO2, as well as the MagnetoHydroDynamics (MHD) effects. The boundary conditions are derived from the results of full banana model, and the results show that the temperature of all materials is not exceeding the upper limits.

Assessment of thermal and radiation induced creep in the dual cooled lead lithium blanket

The creep in recently designed blanket for the fusion nuclear science facility is described by decomposing it into thermal, irradiation and cavity swelling creep. Initial estimates of thermal creep using elastic high temperature rules showed that the recent heat transfer improvements at the first wall (FW) is feasible, and the FW can withstand heat flux of 0.35 MW/m2 at 3000 hrs without rupture. A viscoplastic model that is based on Norton's power law and relies on Multiphysics coupling of solid mechanics and heat transfer modules is used to capture the inelastic deformation in the blanket. Results showed that the design peak heat flux of 0.25 MW/m2 produced maximum thermal creep strain of 0.45 % and the relaxation of the thermal stress at the FW. Irradiation creep is prescribed to be proportional to the displacement damage dose and applied stress. The displacements from irradiation creep radially decrease from the FW to the back wall, but the maximum deformation is found at the back wall that is connected to the Helium manifold due to the high stresses at the region. Cavity swelling creep is dominant at steady state and combined with radiation creep to produce displacement of about 8 mm at the FW.

Design and analyses on the recuperator of the high temperature PbLi experimental platform for CFETR

As a candidate blanket for China Fusion Engineering and Test Reactor (CFETR), the supercritical carbon dioxide (sCO2) cOoled Lithium-Lead (COOL) blanket is under development. To solve the key scientific issue, i.e. MagnetoHydroDynamics (MHD) effect, the high temperature PbLi experimental platform is planned for construction at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). As one of the key components in this platform, the recuperator acts as waste heat recovery and saving electricity power. In this paper, two types of recuperator are designed and analyzed, including the double-tube and hairpin. Firstly, the theoretical and empirical models are used to obtain different designs by changing the structure dimensions and arrangement, then the non-dimensional evaluation criteria is improved and adopted to determine the optimal design that has better comprehensive performance, i.e. higher heat transfer capacity, lower flow resistance to reduce pressure drop and smaller consume volume of liquid metal to increase economy. Finally, the thermal-hydraulic and thermal-mechanical analyses based on the optimal design of the 3D model are further carried out, which obtains the fields of temperature, pressure, velocity as well as stress, and the results can satisfy with the multiple requirements, i.e. outlet temperature under different operating conditions, flow distribution as well as structural strength. The design methodology and analysis results will provide useful technical and data support for the following engineering design and construction of the PbLi experimental platform.

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.

Thermal hydraulic optimization on the First Wall of COOL blanket for CFETR

The supercritical carbon dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket is under development for China Fusion Engineering and Test Reactor (CFETR). The S-CO2 is adopted to cool the structural components, in which the key component First Wall (FW) dominates the total heat power due to the large plasma facing heat flux. In this paper, referring to the response of material temperature, coolant velocity and pressure drop, the thermal hydraulic optimization is firstly performed on the structure design (i.e. cross area and pitch of channel) using the single variable control method, and the optimization principle is that the outlet temperature remains as the same when the optimization variables are changed. Finally, the optimal structure design of FW is obtained, namely cross area of 12 × 12 mm2 and channel pitch of 5 mm. On basis of this, the comparative analyses are carried out using the different cooling system, structural material, i.e. Reduced Activation Ferritic Martensitic (RAFM) steel and Oxide dispersion-strengthened alloy (ODS) ferritic steel, as well as helium as coolant. Furthermore, the satisfied designs are obtained by referring to the constraints of coolant pressure drop and maximum temperature on the structural steel. Then, the thermal hydraulic designs of FW are continuously matching with the Power Conversion System (PCS) to ensure that the heat power removed by the S-CO2 on the primary side can be transferred to the secondary side. Finally, the achievable thermodynamic design are obtained, including the parameters of high operating pressure, turbine inlet temperature, split ratio as well as the thermoelectricity conversion efficiency. The results show that the high pressure of PCS should be larger than 25 MPa when the RAFM steel is used as the structure material, due to the limitation on the outlet temperature of FW. On the contrary, the ODS can raise the outlet temperature, which is beneficial for matching with the PCS, and the thermoelectricity conversion efficiency can reach up to 47.4 % at the high pressure of 25 MPa. The analyses results will provide useful guidance on the system design both for the Primary Heat Transfer System (PHTS) and PCS of CFETR.

Magneto-convective flows around two differentially heated cylinders

Numerical simulations have been carried out in support of an experimental campaign conducted in the MEKKA laboratory at KIT. The aim is investigating liquid metal heat transfer with an imposed magnetic field in a model geometry relevant for the study of water cooled lead lithium blankets for fusion reactors. In the breeding zone of this blanket concept, cooling pipes are immersed in the liquid metal in which convective motion occurs due to significant temperature gradients. The test-section features a rectangular box containing two horizontal cylinders kept at constant differential temperatures in order to establish a temperature gradient that drives the buoyant flow. A magnetic field {textbf{B}} is applied parallel to gravity. The magneto-convective flow, which results from the presence of electromagnetic forces and temperature gradients in the fluid, is relatively complex, since the liquid metal has to move around the cylinders. For weak magnetic fields, a convective recirculation is fed by a jet-like flow formed by the boundary layers that detach from the pipe walls and recombine behind the obstacles. For sufficiently strong textbf{B}, the thermal field resembles that of a conductive regime with vertical isotherms and the fluid is nearly stagnant in most of the cavity except in layers parallel to magnetic field lines and tangent to the cylinders. The rate of convective heat transfer decreases with an increase of the magnetic field. Numerical simulations complement experimental results and give insight into phenomena that cannot be directly analyzed by means of measured quantities.

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&amp;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.

Towards the simulation of MHD flow in an entire WCLL TBM mock-up

In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithium (WCLL) blanket is considered as reference liquid metal blanket design to be tested in ITER and to be used in a DEMO reactor. For the study of pressure and flow distribution in a WCLL blanket module, magnetohydrodynamic (MHD) phenomena, which result from the motion of the electrically conducting breeder in the intense plasma-confining magnetic field, have to be taken into account. They are due to the induction of electric currents that generate strong electromagnetic forces, which modify the velocity distribution in the blanket compared to hydrodynamic conditions and increase pressure losses. In the WCLL test blanket module (TBM) for ITER a basic geometry, consisting of a rectangular box, representing the breeding zone, is repeated along the poloidal direction. The manifold that distributes and collects the liquid metal features two long poloidal ducts, electrically connected across a common wall. The final goal of the present work is to simulate liquid metal MHD flows in strong magnetic fields in a complete column of a WCLL TBM including poloidal manifolds and 8 breeder units. With this aim, the complexity of the model geometry has been progressively increased and various topologies of computational grids have been considered. First simulations have been performed in a single breeder zone connected with a portion of manifolds in order to test the suitability of the generated grid and the performance of the numerical code for this type of problem. In a second step, the MHD flow at moderate magnetic fields in an entire TBM mock-up has been simulated to get pressure distribution along the manifolds and flow partitioning among breeder units.

The Tritium Extraction eXperiment (TEX): A forced convection fusion blanket PbLi loop

A forced convection PbLi loop has been designed and constructed to test tritium extraction from molten PbLi. This Tritium Extraction eXperiment (TEX) is designed to operate at temperatures and tritium concentrations relevant to the helium cooled lead lithium, water cooled lead lithium, and dual coolant lead lithium breeding blankets. The major systems consist of a moving magnet pump, source permeator, supply tank, test section, and plenum. Additional subsystems, including flowmeters, pressure gauges, valves, and heaters are described. The versatile design of TEX will enable code validation using simple tubular membranes, as well as determining tritium extraction efficiencies for alternative test sections.

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.

Steady state thermo-mechanics and material property definition framework for analyzing DCLL blanket in the fusion nuclear science facility

A thermo-mechanics model that relies on creating the material property definition framework (MPDF) and multiphysics coupling of the heat transfer and the solid mechanics modules is developed to determine the structural integrity of the recently designed dual cooled lead lithium (DCLL) inboard blanket (IB) for the Fusion Nuclear Science Facility under steady state loads. The MPDF is called to supply fusion relevant neutron irradiation and temperature induced changes in material properties during multiphysics finite element runs, and PbLi temperature profiles are used to approximate Magnetohydrodynamics effect and the nuclear volumetric heating on the PbLi. Neutron irradiation and temperature induced reduction of the yield and ultimate strengths of F82H steel at the first wall (FW) are quantified for one year. A blanket in an assembly with gaps between blanket sectors and another blanket in an assembly with no gaps between blanket sectors, both exposed to radiation damage that lasted for one year are analyzed. Analysis using the elastic ITER structural design criteria for in-vessel components (ITER SDC-IC) design rules and a linear isotropic-hardening-type elastoplastic material model are used where most appropriate. The IB blanket with gaps between blanket sectors will withstand the steady state combined thermal and coolant loads for one year operational period but will fail if no gaps are allowed between blanket sectors. It is recommended that a gap of about 7.62 mm should be provided between IB blanket sectors during assembly which would close up during service, stop neutron streaming, reduce stresses and reduce bending of the FW into the scrape-off layer.

Neutronics analyses of COOL blanket for CFETR

In this study, an advanced liquid blanket based on a liquid PbLi technology is proposed, and it is called a supercritical CO2 (S-CO2) cooled Lithium-Lead (COOL) blanket. As a candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), it aims to achieve high tritium breeding and shielding performance. It uses S-CO2 as the coolant for the structural components and PbLi as the tritium breeder and neutron multiplier. To evaluate the neutronics performances of this blanket, a detailed 22.5° three-dimensional neutronics model of CFETR was built using the cosVMPT conversion tool. To enhance the shielding capability of the blanket, a shielding block was designed and added at the back of the manifold to moderate neutrons. Evaluation results of the neutron damage of toroidal field coils (TFC) components confirmed that the blanket meets shielding requirements. In addition, the main neutronics parameters including neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heat, and irradiation damage, were calculated using MCNP5. The results indicated that TBR was 1.205, and the total nuclear heat was 1191.64 MW under the fusion power of 1.5 GW. The peak NWL was 1.97 MW/m2, and the maximum DPA and gas production were 14.80 DPA/FPY, 132.65 He appm/FPY, and 631.30 H appm/FPY.

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.

Research on the thermal hydraulic design of COOL blanket for CFETR

The supercritical carbon dioxide (sCO2) cOoled Lithium-Lead (COOL) blanket is under development for Chinese Fusion Engineering and Test Reactor (CFETR). This paper presents the preliminary results of thermal hydraulics on the blanket and shield layer, without considering the MagnetoHydroDynamics (MHD). The mass flow rate in each channel is iteratively adjusted to achieve the outlet temperature of 600 °C and 700 °C, respectively, and the temperature of all materials satisfy with the requirements. The effects of buoyancy-driven at the lower velocity is preliminary observed in the inboard blanket, where occurs the clumps of vortices in the channels. Furthermore, the outlet temperature can further reach 800 °C if the TC is decreased to 2 W/(m·K). At the TC of 28 W/(m·K), the outlet temperature can reach 610.2 °C, respectively, approaching the design edge (600∼700 °C). Moreover, when the TC of co-based cemented tungsten carbide (WC) in the shield layer is 90 W/(m·K), the maximum temperature of WC is 504 °C, and there is a margin of 991 °C far away from the upper limit. It demonstrates that the shield layer can be cooled sufficiently only relying on the manifold.

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.