Dual Coolant Research Articles

Assessment of the possibility of irradiating Tungsten and Cu-alloys in IFMIF-DONES using a realistic specimens configuration

Abstract IFMIF-DONES (International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source) is a facility whose purpose is to obtain data on materials by irradiating them under conditions equivalent to those of the DEMO (DEMOstration) fusion reactor. Tungsten and CuCrZr alloy are two candidate materials to be used in DEMO, tungsten for the first wall and divertor and CuCrZr alloy for the divertor. So, studying the behavior of these materials in IFMIF-DONES facility and comparing the results with the DEMO environment is crucial. In the High Flux Test Module (HFTM) in IFMIF-DONES has been considered a realistic distribution of specimens. The IFMIF-EVEDA beam have been used with a footprint size of 20×5 cm², the nominal energy is 40 MeV with a current of 125 mA, for the calculation other energies have been considered 25, 30 and 35 MeV. The results of IFMIF-DONES have been compared with different DEMO concepts, Dual Coolant Lithium Lead (DCLL), Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB). In general, the primary displacement damage rate meets between IFMIF-DONES and DEMO, while the gas production is at the limit for W and fulfilled for the CuCrZr alloy.

The chemical compatibility of oxide ceramic candidates in static PbLi for advanced breeding blanket channel components

The use of a PbLi eutectic alloy is extended in the Dual Coolant Lead Lithium (DCLL) and Water Coolant Lead Lithium (WCLL) Breeding Blanket (BB) design concepts in future fusion reactors. However, the reactivity of PbLi is a challenging problem, as the flow liquid metal can damage metallic and ceramic materials. For the development of BB technologies, to conduct chemical compatibility studies of materials that will be exposed to the liquid metal is therefore critical. The advanced DCLL concept considers the use of ceramic channels, which could help to extend the BB operating window above 550 °C while improving the tritium-breeding ratio. Since BB components will be highly exposed to neutrons, the influence of PbLi in neutron-like damaged components is also of high concern as compositional changes may modify properties and functionality of the exposed material. Consequently, selected ceramics were submitted to long-term chemical compatibility experiments using COES, a lab-scale facility located at the Liquid Metal Laboratory at CIEMAT, Spain. Commercial Al2O3 ceramics were tested in the as-received and light-ion-implanted conditions, the resultant surfaces after static PbLi exposition being studied by SEM-EDX microstructural and SIMS analytical techniques. The conclusions after the long-term chemical compatibility experiments suggest that the use of bulk alumina ceramics in contact with PbLi is promising. Although lithium diffuses inside the ceramic transforming the initial alumina stequiometry, the surface structural damage is limited to the first 50 µm in the worst case, and could be still considered as a candidate ceramic material for the advanced DCLL BB.

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.

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.

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.

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.

Virtual prototyping of liquid metal blanket performance in fusion pilot plant

Liquid metal blanket is a dominant design option for the next step fusion devices responsible for harvesting energy from fusion reaction, and simultaneously producing fuel for the same reaction through tritium breeding. Liquid metal blankets introduce additional complexity to the design due to fluid motion, fluid structure interaction, and magnetohydrodynamic (MHD) effects arising from the motion of the conducting fluid through the magnetic field. They are also directly affected by the plasma heat flux and neutronic fluence. PPPL is currently developing a virtual prototyping system for numerical analysis of the liquid metal blankets for future fusion devices. The system has a customized 3D computational fluid dynamics (CFD) code in its core, allowing MHD flow and conjugate heat transfer analysis in blankets fluids and solids. The code was successfully used before for dual coolant blanket analysis [A. Khodak et al., Fusion Eng. and Des. 137 (2018)]. Recently the same code was modified to allow verified simulation of MHD flows at high Hartmann numbers of several thousand typical for blanket applications. CFD code receives volumetric heat source distribution from the neutronic analysis based on MCNP code. In addition, direct tritium breeding simulation will be performed allowing optimization of the blanket performance. 2D axisymmetric version of neutronics code will be used for rapid optimization, with 3D version employed for detailed analysis. The surface heat distribution on the plasma facing wall will be defined by the software HEAT allowing 3D modeling of the heat flux based on the magnetic field distribution including gyro-orbit effects. Results of thermal analysis are imported into structural analysis code also included in the system. Direct import of CAD geometry will be used for analyzing all components and as a result design option can be efficiently optimized.

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.

Implementation of CO2 and PbLi as working fluids in RELAP5/MOD3.3 towards accident analysis of COOL blanket for CFETR

The supercritical CO2 cOoled Lithium-Lead (COOL) blanket is an advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). The COOL blanket has dual coolant. Supercritical CO2 (S-CO2) is used to cool the structural components, while lithium-lead (PbLi) is self-cooling in the breeding zones. To verify the reliability of the COOL blanket, it is essential to conduct the accident analysis. The system code RELAP5/MOD3.3 is developed with implementation of the two working fluids of the COOL blanket, namely PbLi and CO2. The outboard blanket segment of the COOL blanket is modeled using the modified RELAP5. The plasma pulse operation and fusion power excursion transient caused by Multifaceted Asymmetric Radiation From the Edge (MARFE) are simulated to study the thermal dynamic behavior of the COOL blanket. Furthermore, sensitivity analysis of Loss of Flow Accident (LOFA) is carried out. The results show that the outlet temperature of PbLi under steady-state operation meets its design requirements, and the temperature of the structural material does not exceed its temperature limit. The thermal hydraulic parameters have pulsed characteristics under plasma pulse operation. The MARFE event can cause sharp temperature rise of the First Wall (FW) but the material remains below its temperature limit. In case of LOFA of the CO2 system, it is suggested to restore CO2 mass flow rate in time to prevent the FW from overheating when the outlet temperature of PbLi is 973 K. The results of the preliminary accident analysis can provide the basis for the design and optimization of the blanket structure and system.

Demonstration of RELAP5-3D for transient analysis of a dual coolant lead lithium fusion blanket concept

Our research aims to build a multiphysics framework for transient analysis of the Dual-Coolant Lead-Lithium (DCLL) blanket design of the proposed Fusion Nuclear Science Facility (FNSF) for all Anticipated Operational Occurrences (AOOs) using Reduced Order Models (ROMs). This paper presents an assessment of using RELAP5-3D for transient thermal-hydraulic blanket analysis of the DCLL design. Preliminary validation studies for the RELAP5-3D properties were conducted using a representative vertical flow loop model and a comparison study using experimental data from thermal convection corrosion loops at ORNL. These studies inspected the thermal hydraulic response of Lead-Lithium Eutectic (PbLi) systems within the RELAP5-3D code; mainly pressure drop and heat transfer, as compared to analytical calculations and experimental data. The Magnetohydrodynamic (MHD) pressure drop effect for liquid metal under the effects of a uniform magnetic field was implemented into our models using an equivalent forms loss coefficient based on the phasic momentum equations. We verified our implementation of MHD pressure drop within RELAP5-3D using a systematic approach consisting of two studies. Models were developed to perform comparison studies using experimental data from the Argonne Liquid metal EXperiment (ALEX) facility and the Magnetohydrodynamic PbLi Experiment (MaPLE) facility. Using the developed validation basis, we built a simplified DCLL channel model within RELAP5-3D, inspected MHD pressure drop and performed transient temperature analysis. The DCLL channel model used radial heating data from the International Thermonuclear Experimental Reactor (ITER) which was normalized to the FNSF design. A representative startup transient was performed using a relative power curve based on current PWR power ramp limits. This study demonstrated that RELAP5-3D ROMs are capable of accurate transient analysis of the DCLL blanket. We have since utilized MCNP to develop heating profiles based on the FNSF design and are working toward implementing them into our DCLL model.

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.

Compatibility of SiC with ODS FeCrAl in flowing Pb-Li at 600°–700 °C

To establish the maximum operating temperature for the dual coolant lead lithium (DCLL) blanket concept, thermal convection loop (TCL) experiments were conducted for 1000 h with a peak temperature of 700 °C to determine compatibility with commercial eutectic Pb∼17at.%Li. Two TCLs were fabricated from alloy APMT (FeCrAlMo) and the baseline TCL experiment contained only APMT specimens. The second TCL experiment exposed more fusion relevant CVD SiC and ODS Fe-10Cr-6Al specimens in the hot and cold legs. In the second experiment, the mass losses were much larger for the ODS FeCrAl specimens and most of the CVD SiC specimens experienced a mass gain as SiC reacted to form an Fe- and Cr-rich carbide reaction product. Dissolved Fe and Cr reacted with the SiC and removing Fe and Cr from the liquid appeared to increase ODS FeCrAl dissolution. Such a dissimilar material interaction appears consistent with thermodynamic calculations. Significant attack at the top of the hot leg in both experiments suggests that the Pb-Li compatibility of FeCrAl alloys is inadequate at 700 °C in a flowing environment and the maximum PbLi temperature is < 675 °C.

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.

Three-dimensional magneto convective flows in geometries relevant for DCLL blankets

Abstract The dual coolant lead lithium (DCLL) blanket is one of the liquid metal blanket concepts investigated within the framework of the EUROfusion Consortium. In the proposed design, the liquid metal serves as tritium breeder, neutron multiplier, and as coolant for the breeding zone. Magnetohydrodynamic (MHD) effects resulting from the interaction of the moving liquid metal with the magnetic field that confines the fusion plasma are coupled with thermal phenomena. Induced electromagnetic forces lead to significant modifications of the hydrodynamic flow and affect in turn heat and mass transfer. The presence of a high-intensity neutron flux in the liquid metal yields volumetric heating and associated buoyant forces that drive complex flow patterns in the blanket. The resulting convective motions cause strong thermal mixing, and an increment in the effective heat transfer coefficient that may increase heat losses from the liquid metal into the helium coolant. This could deteriorate the thermal efficiency of the blanket. Buoyant MHD phenomena may also result in reversed flow and closed recirculations where tritium accumulates and temperature locally increases. The aim of the present investigation is to address by means of 3D numerical simulations the main features of magneto-convective flows in a geometry typical for DCLL blankets, in which PbLi moves upwards in a poloidal duct that faces the first wall and, after making a U-turn, flows downwards at the back of the blanket module. The liquid metal enters and leaves the model geometry by flowing radially in rectangular channels. It is found that in the front duct the velocity is characterized by jets in the boundary layers along walls parallel to the magnetic field, as expected in case of electrically conducting ducts. In the rear channel, along the wall that separates the two poloidal ducts, a downward flow occurs with much higher velocities than in the rest of the channel. The influence of the magnetic field on these flow features is investigated.