Blanket Coolant Research Articles

3D MHD analysis of prototypical manifold for liquid metal blankets

The water-cooled lead lithium (WCLL) and the dual-cooled lead lithium (DCLL) are two of the breeding blanket (BB) concepts that the EUROfusion consortium is pursuing in the framework of the development of the fusion reactor industrial demonstrator DEMO. Both involve the use of a liquid metal (LM) as working fluid, the lead-lithium eutectic alloy (PbLi), due to its excellent thermal properties and the possibility to serve as both the blanket coolant and tritium breeder and carrier. Unfortunately, due to the high electrical conductivity of LMs, their motion is influenced by the magnetic field used in the reactor to confine the plasma, generating a complex phenomenology which is studied by magnetohydrodynamics (MHD). In this work, a representative prototypical manifold of a BB bottom feeder is investigated for different configurations with the custom phiFoam solver, capable of simulating unsteady, incompressible and isothermal MHD flow. The aim of this study is to investigate which configuration minimizes the flow imbalance in the manifold for the WCLL or in the poloidal breeding zone channels for the DCLL. The distribution of the flow rate among the channels is strongly influenced by the position of the feeding pipe (FP) and by the development of the MHD internal layer near the expansion, which generates important jets close to the lower plate and the upper one, where the channels are attached. The channel aligned with the FP is the one carrying most of the flow, from 55% to 82%, while in the more distant one the flow is almost stagnant, carrying from 17% to 6% of the total flow rate. The total pressure loss is also estimated and its functional dependence on the manifold configuration is discussed.

3D Transient CFD Simulation of an In-Vessel Loss-of-Coolant Accident in the EU DEMO WCLL Breeding Blanket

The in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case.

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.

Effect of different system parameters on the design of the EU DEMO Vacuum Vessel Pressure Suppression System

The Vacuum Vessel Pressure Suppression System (VVPSS) is the safety system devoted to the mitigation of in-vessel Loss-Of-Coolant Accidents, which are considered among the Design-Basis Accidents for tokamak fusion reactors. The EUROfusion Consortium is performing the design of this system, as part of the conceptual design phase of the EU DEMO fusion reactor, including in the process its integration in the plant.This work reports the first parametric calculations carried out to evaluate the VVPSS response to a postulated double-ended guillotine rupture of a Breeding Blanket coolant feeder. The aim of the study is to characterize the role that key parameters play on the multiple systems involved in such an accidental event. Assessments scanned a broad range of design variables, some closely related to the suppression system itself, and others dependent on Primary Heat Transfer System features. The parameter space is mapped identifying the regions where the accident consequences are fully mitigated, in terms of peak pressure in the Vacuum Vessel (VV) and equilibrium pressure of both VV and VVPSS, providing a design tool to rapidly adapt the VVPSS to the evolving plant design.

Power conversion from spherical tokamak test reactor with helium-cooled and water-cooled blanket

Possible configurations for Power Conversion System (PCS) for a futuristic, small, pulsed Spherical Tokamak (ST) fusion reactor are presented. For the typical blanket coolant choices namely, helium and water, an intermediate heat-exchanger (IHX) cum buffer energy storage system (ESS) with Molten Salt (MS) is introduced to examine Rankine and Brayton cycle’s effectiveness. Preliminary thermal sizing of the involved heat-exchangers and parameters of the optimized cycle are discussed. S-CO2 Brayton cycle seems to be a promising candidate for efficient power conversion for a pulsed plasma reactor with efficiency of 42.75%. Rankine cycle yields efficiency of 35–42% depending on water/helium as the choice of the blanket coolant.

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

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

Study of the corrosion characteristics of 304 and 316L stainless steel in the static liquid lithium

• Corrosion behavior of 316L SS and 304 SS exposed to liquid Li at 600 K for 1320 h under Argon (Ar) atmosphere were investigated. • The corrosion characteristics of 316L SS and 304 SS was compared. • 316L SS was found to be more corrosion resistant in high-temperature liquid lithium than 304 SS. Liquid lithium (Li) is considered as the tritium breeder as well as a potential blanket coolant and/or plasma-facing material (PFM) in fusion devices, while steels are widely used as an important structural material in the fusion reactors. Investigation of corrosion behaviors of steels exposed to liquid Li is one of the keys to simultaneous apply Li and steels as potential PFM or blanket materials in the fusion devices. Hence, in this study, the corrosion behavior of 304 stainless steel (304 SS) and 316L stainless steel (316L SS) in static liquid Li, at 600 K, for 1320 h, under Argon (Ar) atmosphere is analyzed. After exposed to liquid Li, the mass loss of 304 SS is about 3.6 times more than that of 316L SS. And the corrosion depth rate of 316L SS and 304 SS is 0.71 μ m / a and 2.6 μ m / a , respectively. Both the materials have shown a non-uniform corrosion behavior. The Signs of grain boundary attack, pitting corrosion, and corrosion products formation, due to the selective dissolution of elements and chemical reactions are also observed. The percentage of corrosion product formation (i.e., carbides) on the surface of 304 SS, is higher than 316L SS. The hardness of two SS materials has a slightly increased after liquid Li exposure.

Numerical investigation of the tritium permeation phenomenon through cooling plates in breeding blankets

Understanding and predicting tritium transport is considered a key issue for the design of safe and self-sufficient fusion power plants. Indeed, tritium generated in the breeder materials is susceptible to permeate to the blanket coolant circuits where it is harder to remove and process. Therefore, the predictive models become very important to evaluate the performance of any blanket design. In this work, 2D finite volume methods are employed to analyze in detail the tritium permeation through a cooling plate considering a set of different parameters that affect the breeding blanket (BB) performance. Different BB configurations are studied varying the geometrical parameters defining the plate: the wall thickness and the cooling channel pitch. Moreover, the analyses cover every permeation regime, from diffusion-limited to surface-limited, including intermediate situations. This way, results are applicable to every BB concept which employs cooling plates with independence of their specific materials, temperatures and tritium concentrations. Results have been compared with those of 1D models in order to define form factors that can be used to increase the precision of system level tritium transport models. A total of 91 simulations have been carried out to find correlations that allows computing the form factors in a wide variety of situations. Results show that the accuracy of 1D models and consequently the accuracy of most system level models strongly depends on the permeation regime and can be poor in some conditions. However, this study demonstrates that the use of form factors can be used for an effective increase of system level exactness.

Status of the DEMO blanket attachment system and remaining challenges

The DEMO blanket attachment concept is challenging due to several factors: the harsh radiation environment, the thermal expansion, the electro-magnetic loads, the remote maintenance feasibility, and the accurate control of the alignment of the breeding blanket first wall during operation. There are two inboard and three outboard blanket segments per vacuum vessel sector to be installed and extracted by remotely controlled tools through a single upper vertical port. The design of the fixations of the blanket segments to the VV complies with the strategy to avoid the need for front side access for engagement and release. The attachment system has been designed for the numerous critical load cases, including normal operation, dwell between pulses, plasma disruptions, fast discharge of the magnet coils, and accidental conditions such as loss of blanket coolant. At the same time the attachments must guarantee the stresses in the blanket segments not to exceed limits.This paper introduces the attachment concept and describes the finite element model that has been built to assess the blanket attachment system. The model represents one sector of the DEMO machine. The results focus on the reaction forces transmitted at individual attachment locations to define these interfaces and guide the design of the individual supports.

On the thermal dynamic behaviour of the helium-cooled DEMO fusion reactor

The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to mitigate the potential negative impact of plasma pulsing on BoP equipment, for the DEMO plant is also being investigated a “heat transfer chain” option which envisages an Intermediate Heat Transfer System (IHTS) equipped with an Energy Storage System (ESS) between PHTS and PCS. Within this framework, a preliminary study has been carried out to analyse the thermal dynamic behaviour of the IHTS system for the Helium-Cooled Pebble Bed (HCPB) BB concept during pulse/dwell transition which should be still considered as the normal operating mode of a fusion power plant. Starting from preliminary thermal-hydraulic calculations made in order to size the main BoP components, the global performances of DEMO BoP have been quantitatively assessed focusing the attention on the attitude of the whole IHTS to smooth the sudden power variations which come from the plasma. The paper describes criteria and rationale followed to develop a numerical model which manages to simulate simple transient scenarios of DEMO BoP. Results of numerical simulations are presented and critically discussed in order to point out the main issues that DEMO BoP has to overcome to achieve a viable electricity power output.

Scaling analysis on LOFA for the test model of WCCB blanket in CFETR

The water Cooled Ceramic Breeder (WCCB) blanket, as a candidate for Chinese Fusion Engineering Test Reactor (CFETR), must demonstrate its safety before engineering application. Enough data are necessary and the test model should ensure that the data are convincing within reasonable expenses. In this paper, scaling criteria are identified for loss of flow accident (LOFA) in WCCB blanket coolant system, which guides the design scheme of test model that the length scaling factor of 1 and volume scaling factor of 1/120. Scaling analyses are performed to develop similarity criteria for pump coastdown and loop natural circulation in three temporal phases identified by LOFA simulation. The test model only preserves a branch at the sector level and sub-module level, which results in different volume scaling factors at levels. Typical heat-transferring cells, constituted by blanket components, are preserved for heat transfer similarity. By computational simulation of an ideal model, it is testified that the test model can reproduce thermal-hydraulic responses within acceptable range of the distortion.

Investigation of wettability of Li on 316L SS surface and interfacial interactions for fusion device

Improved liquid Li wettability on a solid substrate surface is a necessary component for the research and development of blanket coolants, tritium breeders, and first wall materials for fusion device. A systemic investigation of Li wetting a 316 L SS surface, along with interfacial interactions has been successfully carried out. The results indicate that base vacuum pressure of the vessel, substrate temperature, and surface conditions of the substrate are the main factors that influence Li wetting property. It has been found that poor vacuum (>10−1Pa) along with high impurity content, especially water, can prevent Li wetting due to the formation of a layer of complex Li compounds, such as LiOH, Li3N and Li2CO3. Also, Li wetting can be gradually improved with an increase of substrate temperature, mainly due to a decrease in surface tension of Li droplets. Moreover, Li wetting can be further improved by effective surface conditioning. It has been demonstrated that baking was beneficial in order to release impurity gas by thermal desorption and thereby achieve high vacuum during Li wetting, which resulted in a reduction of > 50 °C in Li wetting temperature on the SS substrate. He-GDC and Li coating can further decrease Li wetting temperatures to the Li melting point (∼180 °C) by avoiding the formation of Li2O at the interfacial surface during Li wetting. This is achieved primarily because these conditioning techniques can effectively remove and isolate the SS surface oxide layer (Cr2O3), respectively, which was verified by the related testing of interfacial interactions between liquid Li and SS. Finally, surface textured structures with trenches ∼μm in size, are demonstrated to promote Li spreading along the groove direction, driven by capillary forces. These results provide the technical support for liquid Li applications in future fusion reactors.

Corrosion characteristics of copper in static liquid lithium under high vacuum

Copper (Cu) materials are extensively used as heat sinks and sealing gaskets in fusion devices because they have the properties of good thermal and electrical conductivity and high plasticity. Meanwhile, liquid lithium (Li) is considered as a potential blanket coolant and tritium breeder and/or plasma facing material in fusion devices. Studying the corrosion characteristics of Cu materials by liquid Li under extreme fusion conditions is important because the corrosion of Cu by liquid Li may affect simultaneous application of these materials in fusion devices. The corrosion behavior of Cu in static liquid Li at 620 K for 15 h under high vacuum was investigated. After exposure to liquid Li, the weight loss rate of Cu in liquid Li is 466.1 g m−2 h−1, which is equivalent to 458.7 mm⋅a−1 for the average corrosion depth rate. The entire surface of each specimen was seriously damaged. Visible grain boundary corrosion was observed on the surface of the specimens. Also, Cu debris entered the liquid Li from the corroded surface, resulting in considerable Cu loss from the specimen. These results demonstrate a corrosion protection grade of Cu in liquid Li of 10, Cu cannot withstand the corrosion of liquid Li under the given conditions. Additionally, the corrosion process of Cu in liquid Li at 620 K under high vacuum was studied using the Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS). The results from these simulations indicate that the corrosion of Cu in liquid Li is induced via physical dissolution and intergranular corrosion, where intergranular corrosion is the dominant mechanism.

Sizing of a recuperative supercritical CO2 Brayton cycle as power conversion system for DEMO fusion reactor based on Dual Coolant Lithium Lead blanket

A modified recuperative supercritical CO2 cycle for DEMO fusion reactor is analysed. A layout with just one recuperator is used, because of the relatively low temperature of the breeding blanket coolants. The proposed cycle is able to integrate several thermal sources with different temperature ranges and includes two kinds of thermal energy storage to overcome the pulse operation: a two-tank molten salt storage and a concrete storage. The design point is obtained using the Engineering Equation Solver (EES) environment, which allows to use pure substance models for CO2, helium and water. Once the design point is calculated, the sizing of pipes, heat exchangers, turbomachines and energy storage systems has been performed in order to obtain a preliminary estimation of the layout and its complexity. So, the power conversion system has been split into four equal parts in order to use commercial pipe diameters. Even though Printed Circuit Heat Exchangers (PCHE) have been used, extremely large sizes have been obtained, especially for the HTR recuperator. These sizes contrast with the small values obtained for the turbomachines, whose preliminary design has been carried out, using the AXIAL™ meanline design software. In addition, the two largest heat exchangers are divided into serial and parallel configurations when assessing the allocation of the components of the power conversion system to reduce the footprint. An average electric efficiency (including pulse and dwell period) of 28.25% has been reached. The required mass of molten salt per each power conversion subsystem is 2568 t, while the mass of concrete is 833 t. Regarding the footprint, 85 × 170 m2 are required for each power conversion subsystem, including the storage systems.

Preliminary design of EU DEMO helium-cooled breeding blanket primary heat transfer system

The European DEMO conceptual design foresees four Breeding Blanket (BB) concepts that rely on different cooling and breeding technologies. As DEMO has been conceived to deliver net electricity to the grid, the choice of the blanket coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. Moreover, the machine pulsed operation makes the BB Primary Heat Transfer System (PHTS) the main hub of the DEMO Balance of Plant (BoP).Within this framework, a study has been carried out to design the Ex-Vessel PHTS of the Helium-Cooled Pebble Bed (HCPB) BB concept.The paper describes criteria and rationale followed with the aim to achieve a simple PHTS design based on the adoption of easy-to-manufacture main components. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, piping and compressors are presented and critically discussed. Finally, the evaluation of PHTS key parameters as total pressure drops and total coolant inventory is reported to provide input relevant to the assessment of the impact of helium technology on the design integration and feasibility of DEMO BoP.

Study of the corrosion behaviors of 304 austenite stainless steel specimens exposed to static liquid lithium at 600 K

Investigation of corrosion behavior of stainless steel served as one kind of structure materials exposed to liquid lithium (Li) is one of the keys to apply liquid Li as potential plasma facing materials (PFM) or blanket coolant in the fusion device. Corrosion experiments of 304 austenite stainless steel (304 SS) were carried out in static liquid Li at 600 K and up to1584 h at high vacuum with pressure less than 4 × 10−4 Pa. After exposure to liquid Li, it was found that the weight of 304 SS slightly decreased with weight loss rate of 5.7 × 10−4 g/m2/h and surface hardness increased by about 50 HV. Lots of spinel-like grains and holes were observed on the surface of specimens measured by SEM. By further EDS, XRD and metallographic analyzing, it was confirmed that the main compositions of spinel-like grains were M23C6 carbides, and 304 SS produced a non-uniform corrosion behavior by preferential grain boundary attack, possibly due to the easy formation of M23C6 carbides and/or formation of Li compound at grain boundaries.

Development of Sandwich Flow Channel Inserts for an EU DEMO Dual Coolant Blanket Concept

The dual-coolant lead-lithium (DCLL) blanket concept, which is considered as a candidate for fusion power plants and possibly for a demonstration reactor (DEMO), is being investigated within the framework of the European Power Plant Physics and Technology (PPPT) study. One of major issues of the DCLL concept development is the design of the flow channel inserts (FCIs), which are essential for the reduction of magneto-hydrodynamic (MHD) pressure losses. Due to the tight schedule for the short-term PPPT DEMO, a low-temperature DCLL concept with a liquid metal outlet temperature below 500 °C has been proposed. This allows the use of a simpler type of FCI (taking into account the LM corrosion issues), e.g. Eurofer-Alumina-Eurofer sandwich FCI, instead of the SiCf/SiC version for high temperature case, the production thereof is challenging. This paper discusses the technological study on manufacturing of some FCI design variants and post-examination of the samples.

Magnetohydrodynamic flow in ducts with discontinuous electrical insulation

In liquid metal blankets the interaction of the moving breeder with the intense magnetic field that confines the fusion plasma results in significant modifications of the velocity distribution and increased pressure drop compared to hydrodynamic flows. Those changes are due to the occurrence of electromagnetic forces that slow down the core flow and which are balanced by large driving pressure heads. The resulting magnetohydrodynamic (MHD) pressure losses are proportional to the electric current density induced in the fluid and they can be reduced by electrically decoupling the wall from the liquid metal. For applications to dual coolant blankets it is foreseen to loosely insert electrically insulating liners into the ducts. In long channels the insulation could consist of a number of shorter inserts, which implies a possible local interruption of the insulation. Three dimensional numerical simulations have been performed to investigate MHD flows in electrically well-conducting channels with internal discontinuous insulating inserts. The local jump in the electric conductivity of the duct wall results in induced 3D electric currents and related electromagnetic forces yielding additional pressure losses and increased velocity in boundary layers parallel to the magnetic field.

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

The findings of the EU ‘Materials Assessment Group’ (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision.A DEMO phase I with a ‘Starter Blanket’ and ‘Starter Divertor’ is foreseen: the blanket being capable of withstanding ⩾2MWyrm−2 fusion neutron fluence (∼20dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5MWyrm−2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650°C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290–320°C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200–350°C) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds.Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed.Testing with 14MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ⩾30dpa (steels) 14MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14MeV testing programme requirements.

Liquid Metal MHD Flows Near Noninsulated Gaps Between Flow Channel Inserts in DCLL Blankets

Dual coolant liquid metal blankets with helium cooled structure and self-cooled breeding zone have been proposed for a DEMO reactor. The feasibility of this concept is strictly related to the possibility of reducing the pressure drop caused by the intense magnetohydrodynamic (MHD) interactions, namely related to electromagnetic forces induced in the fast flowing liquid metal. In order to minimize the MHD pressure losses, electrically insulating channels should be used. The interruption of the current path through conducting walls could be reached by fitting low conducting flow channel inserts into the ducts. Due to fabrication issues or maintenance requirements, the insulating liners could present axial gaps, which determine a local discontinuity of the electric insulation at the wall. The MHD flows close to gaps of insulating inserts in electrically conducting channels are investigated by asymptotic methods and by 3 -D numerical simulations for different magnetic fields and various values of the electrical conductivity of the walls. The additional pressure drop resulting from 3-D current loops that close through pipe walls and in the fluid is quantified.