Breeding Blanket Research Articles

Supporting analysis for WCLL test blanket system safety

The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

Manufacturing of W-steel joint using plasma sprayed graded W/steel-interlayer with current assisted diffusion bonding

The differences in the thermophysical properties between tungsten and steel create thermal stress peaks at its interface when joined directly for the breeding blanket of a fusion reactor. In order to solve this problem, a graded interlayer made of several layers of W/steel-composites is considered to reduce these stress peaks. Plasma spraying under an argon shrouded chamber was employed as a cost efficient manufacturing technique for such composites and field assisted sintering technology/spark plasma sintering was used to diffusion bond them with bulk-W and bulk-steel. Firstly, thermophysical characterizations were performed on these composites. Secondly, two approaches have been investigated to join bulk-W and 75 vol% W composite: direct joining and using a vanadium foil. Only vanadium foil resulted in successful bond formation at all the three bonding temperatures of 800 °C, 900 °C and 1000 °C. Thirdly, investigation of diffusion bonding parameters (temperature and time) for the joining of 25 vol% W, 50 vol% W, 75 vol% W and bulk-steel were studied and optimum process parameter were identified. Finally, this optimized parameter (1000 °C; 30 min) was employed to manufacture a complete 12 mm x 12 mm W-steel joint consisting of this graded interlayer.

Static structure, collective dynamics and transport coefficients in the liquid Li-Pb alloy. An ab initio molecular dynamics study

Several static and dynamic properties of the liquid Li-Pb alloy at diverse compositions, have been calculated by means of ab initio molecular dynamics simulation study. This alloy has attracted much attention because of the finding of fast sound at the Li0.80Pb0.20 composition and also the technological interest of the the eutectic composition, Li0.17Pb0.83, as a component of the blanket in fusion reactors.Results are reported for total static structure factors, which are compared with the available experimental data. An additional analysis of the structure allows the quantification of the heterocoordinating tendencies in this alloy, which at the Li0.80Pb0.20 composition are largest and lead to a Pb-centered polyhedral structure, where, however, Li4Pb units are not present.Regarding the collective dynamics, the calculated partial dynamic structure factors exhibit side peaks indicative of propagating density fluctuations, including density fluctuation modes with phase velocity greater than the hydrodynamic sound velocity. Also, the longitudinal and transverse dispersion relations have been calculated and its different branches analysed. We find all the high frequency branches to behave as optic-like modes, contrary to other interpretations in terms of an acoustic-like fast sound mode.Some transport coefficients such as self- and inter-diffusion coefficients, shear viscosities and adiabatic sound velocities, have also been calculated. Finally, the obtained results for the electronic density of states clearly indicate the metallic character of the liquid LixPb1-x alloy.

Design and integration studies of a diagnostics slim cassette concept for DEMO

The diagnostics slim cassette (DSC) is a concept currently proposed for DEMO to host the microwave reflectometry diagnostic in a dedicated poloidal section, for the feedback control of plasma position and shape, also providing the radial edge density profile at several poloidal angles. This DSC, which is expected to house up to 100 antennas and waveguides, is to be integrated with the breeding blanket (BB) segments. Moreover, it is being designed with remote handling compatibility in mind to facilitate a ‘fast’ exchange by remote maintenance (RM) when needed, fulfiling one of the aims of the work package diagnostic and control. In this approach the pre-assembled (banana-shaped) DSC modules are inserted/removed from the upper port (UP) of the vacuum vessel (VV), at least when the BB segments are replaced. Here the main constraints related to the integration of the DSC with the BB, the UP and the equatorial port are identified and discussed in detail, with a strong focus on the required RM operations inside the VV (in-vessel), providing solutions that can be adapted to present and future blanket and UP designs.

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.

Assessment of residual heat removal from activated breeding blanket segment during remote handling in DEMO

In this study, the temperature of the activated EU DEMO breeding blanket (BB) during the in-vessel remote handling is analyzed considering the decay heat generated inside the BB. A transient analysis of the natural convection of the injected air inside the plasma chamber has been performed using the Computational Fluid Dynamics (CFD) simulations. The CFD results are exploited for tuning of a lumped model, which has been used for fast predictions of longer transients. The aim of this study is to determine the temperature and temperature gradient of the BB during the remote handling, the temperature of the circulating air as environmental condition for the RH tools, and to investigate the feasibility of residual heat removal from BB segments by natural convection of filled gas. The removal of the BB segments by remote handling (RH) is foreseen here one month after the plasma shutdown.

Activation Analysis and Evaluation of Radionuclide Inventory Decay Heat for EU DEMO Vacuum Vessel Components

Activation inventory data were assessed for different European DEMOnstration Fusion Power Reactor (EU DEMO) vacuum vessel (VV) segments in order to evaluate the activity, dose rates, and decay heat of the structural materials and the impact of heterogeneity as a part of the Safety and Environment project of the EUROfusion/Power Plant Physics Technology program. Activity is a good indicator of the condition of irradiated material while decay heat and dose rate estimates are crucial for power plant operation and public safety. In addition, the VV structures were analyzed in terms of different breeder blanket configurations that were implemented in the model. Helium-cooled pebble bed and water-cooled lithium-lead breeder blanket modules were considered. These modules directly shield the VV from neutron irradiation; therefore, their design has a great influence on subsequent structure activation.

Investigation on the Tritium Self-Sufficiency and Neutron Shielding Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a new test tokamak device in China, which is to bridge the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.05. In addition, the neutron shielding evaluation is also a significant issue, which could provide a significant reference for the water-cooled ceramic breeder (WCCB) blanket modification and improvements. In this work, a 3-D simplified neutronics model containing the WCCB blanket is developed and the neutronics performance is calculated based on the 1.5 GW fusion power. The nuclear analyses are carried out by the Monte Carlo N-particle (MCNP) transport code, including the TBR, neutron wall loading (NWL), fast neutron flux on inboard side, and nuclear heating deposition on main in-vessel components. It is found that the net TBR reaches 1.23, which meets the condition of tritium self-sufficiency. Moreover, it is found that the peak NWL of inboard and outboard occurs in the equatorial plane. The inboard peak NWL is 1.89 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> and the outboard peak NWL is 1.97 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . In addition, the fast neutron flux and nuclear heating deposition at the inboard midplane in toroidal field coil (TFC) are obtained. According to the radiation limits of the CFETR TFC referring to ITER, the results show the nuclear responses of the TFC meet the limit criteria.

The Compactness and Inboard Radial Build of Fusion Nuclear Devices

Restraining the size of fusion power plants is considered an important avenue to make them a competitive energy source among other forms of energy production. The most critical contributor to the size of a tokamak is the inboard radial build, composed of multiple components with various functions. This build is the ultimate limit to size reduction. The Fusion Nuclear Science Facility is reviewed and each element of the inboard build is described, showing that the build, including breeding blanket, structural ring, vacuum vessel, low-temperature shield, and toroidal field and central solenoid (CS) coils, contributes 2.9 m of build, with 0.6 m of bore hole inside the CS coil, or 3.5 m to reach the plasma scrape-off layer. This implies that it would be challenging to make a significantly smaller build and simultaneously meet all the engineering requirements.

Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods.

Modeling Magnetic Material Using Vector Potential-Based Elements in ANSYS and Comparing with Edge-Based Elements

More and more, soft or hard magnetic materials have been used in the magnetic confinement fusion machine, like reduced-activation ferritic-martensitic steel, which may be used in the first wall and breeding blanket structural material of ITER. Twisted coils in the stellarator may be substituted by permanent magnets. ANSYS electromagnetic simulation has two methods: nodal-based formulation or edge-based formulation. Nodal-based formulation has been widely used in past years. This paper will present a modification to the nodal-based element SOLID97 in ANSYS to form the correct normal boundary condition to model ferromagnetic materials and compare the results, e.g., field, eddy current density, and Lorentz force, to edge-based methods using SOLID236.

Effects of He irradiation on the microstructure and mechanical performance of Li2TiO3

In solid fusion blankets, energetic helium ions produced by the breeding reaction bombard Li2TiO3 and influence its microstructure and mechanical performance. Here, evaluations of the microstructure and mechanical performance are performed for Li2TiO3 irradiated with 2 MeV helium ions. The Raman spectra demonstrate irradiation-induced structural disorders, which are corroborated by transmission electron microscopy micrographs and selected area electron dispersion patterns. Nanoindentation investigations indicate that the hardness is influenced by irradiation damage. At low doses (about 0.018 dpa), the decrease in hardness is caused by dislocation motion and a decrease of the bond density per unit volume. At high doses (about 0.86 dpa), the defect clusters formed impede the mobility of dislocations, which is the main reason for the increase in hardness. Based on hardness theories and Raman spectra, the decrease in bond length is considered to be another reason for the increase in hardness.

Measurement of effective thermal conductivity of lithium metatitanate pebble beds by steady-state radial heat flow method

As a functional material of the tritium breeding blankets of the future fusion reactor, different candidates of the lithium-based ceramics have been considered to generate and release tritium. These tritium breeding materials in the form of packed pebble beds of nearly spherical-shaped particles have many advantages over the sintered pellets or blocks. For the design and analysis of the breeding blanket, the effective thermal conductivity (keff) of lithium-based ceramic pebble beds are required to be well characterized under the fusion-relevant conditions. In this study, an experimental setup has been fabricated and installed to estimate the keff of pebble beds as a function of bed temperature and filling gas pressure. The working principle of the experiment is based on the steady-state radial heat flow method. The various experiments have been performed on lithium metatitanate (Li2TiO3) pebbles having a diameter of 1 ± 0.15 mm to measure keff in the temperature range from 400 °C to 800 °C and in the environments of stagnant helium gas. The pressure of helium gas was adjusted from 0.105 MPa to 0.4 MPa (abs.) to account the effect of helium gas pressure. The increase in keff has been found with increasing the bed temperature and also increasing the helium gas pressure. The obtained results have been found comparable with the literature results with minimum difference.

The effects of side interactions on the neutronic analysis of HCCR breeding blanket

Helium Cooled Ceramic Reflector (HCCR) is a candidate as Multi-Module Segmentation Breeding Blanket used the Korean DEMO Reactor (K-DEMO). Neutronic analyses of the breeding blanket are substantial due to the prominent role of neutrons in tritium breeding, energy deposition, and radiation damages. This work aims to consider the effects of side reactions with low cross-sections on the neutronic study to make the simulation results more accurate. The neutronic analysis of the HCCR blanket has been performed using the MCNPX.2.6.0 3D Monte Carlo code and ENDF/B-VII nuclear data library. The effects of side reactions have been considered on Tritium Breeding Ratio (TBR) and Neutron Multiplication Factor (NMF) computations. The 9Be(n,T)7Li is a side reaction to tritium breeding, and the 7Li(n,2n) and W(n,2n) reactions with low cross-section cause neutron multiplication. Furthermore, the deposited energy by alpha particles has been investigated. Finally, the reaction rate of the photonuclear response has been examined using MCNPX.

Simulation of loss of off-site power accident of CN HCCB TBS

Helium Cooled Ceramic Breeder Test Blanket System (HCCB TBS) is proposed by China, which will be tested in ITER facility to validate the tritium breeding blanket technology. Test Blanket Module (TBM) is the core component associated with its helium cooling system (HCS) and other ancillary systems. The accident analysis is necessary during the preliminary design process, and the results obtained provide safety feedback to the following design work. Loss of Off-site Power (LOSP) is one of the design basis accidents (DBA), and is simulated using the best estimated code RELAP5 MOD3.4. For the safety simulation, besides the thermal conductive and convective model, the radiation model is adopted as well. The influence of the natural circulation of the HCS primary loop and the necessity of the Class III power supply (safety class) are discussed. The simulation results show that the peak temperature of the structure material exceeds the design temperature (550 °C) for short duration, which has almost no impact on the structure. In the meantime, the duration of the temperature over 550 °C drops sharply while the natural circulation and the Class III Power are considered. The results verify the integrity and the safety of the HCS primary loop during the LOSP accident.

Mechanical support concept of the DEMO breeding blanket

• Description of blanket integration concept. • Stress calculation of breeding blanket segments. • Prediction of Breeding blanket deformation. • Identification of custom-machining concept and prediction of achievable tolerances. • Overview of the various integration issues associated with the design of in-vessel components. The DEMO tokamak architecture is based on large vertical breeding blanket (BB) segments that are accessed from a maintenance hall above the tokamak and are vertically replaced through large upper ports of the vacuum vessel (VV). The feasibility of the BB segments mechanical supports is a prerequisite of this vertical segment architecture. Their design directly impacts on the removal kinematics and the remote handling operations required for release and engagement. The supports must withstand large forces acting on the BB in particular due to electromagnetic (EM) loads. At the same time, they must ensure a sufficiently precise positioning of the BB first wall. Their design also takes into account the significant thermal expansion of the blanket segments that are operated at high temperature avoiding excessive support reaction forces. The BB support concept described in this article does not require fasteners or electrical straps to the VV and therefore much reduces the complexity of the BB remote replacement – a valuable characteristic that would make this concept a milestone in meeting one of the goals defined for the DEMO project: to develop a maintainable fusion power plant design [1] . Each blanket segment is individually supported by the VV without any physical contact to the other blankets or in-vessel components. It relies instead on vertical pre-compression inside the VV due to obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the BB material in the toroidal magnetic field. The verification process did not identify show stoppers. Nonetheless, a further evolution of the concept is required including design improvements to mitigate the high stress levels found in the inboard blankets during plasma disruptions. The fact that no excessively high support reaction forces or large BB deflections were found suggests though that the further development of the concept could be successful.

Corrosion of reduced activation ferritic-martensitic steel CLF-1 by Li2TiO3 in ambient air

In a solid tritium breeding blanket, tritium breeder Li2TiO3 is in contact with reduced activation ferritic martensitic steel at high temperatures for years and accordingly form a corrosion layer on the blanket structural steel. The effects of the corrosion layer on the deuterium permeability of RAFM steel CLF-1 were thus investigated. X-ray diffraction, secondary ion mass spectroscopy and high-resolution transmission electron microscopy proved the formation of a double layered corrosion products with a thickness of ∼200nm . A localized LiCrO2 (003)/LiFeO2 (003) interface between the outer layer LiFeO2 and the inner layer LiCrO2 was observed. With the increase of corrosion periods (120∼360 h), there is a first deuterium permeability increase due to cracking and an amorphization process occurred at the interface between the corrosion layer and the substrate. The CLF-1 steel corroded for 720 hours performed a lowest deuterium permeation flux as the amorphous oxides have been entirely crystallized. A localized corrosion mechanism coexisting with uniform corrosion was proposed to explain the changing deuterium permeability of the CLF-1 steel corroded by Li2TiO3 in 1 atm ambient air atmosphere. Our study provides a novel thought on reducing the tritium permeability of the RAFM steel.

Model Development and Transient Analysis of the WCLL BB BOP DEMO Configuration Using the Apros System Code

Extensive modelling and analytical work has been carried out considering the water-cooled lithium–lead breeding blanket (WCLL BB) balance of plant (BOP) configuration of the demonstration power plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. Contributing to the BOP work package of the EUROfusion Consortium, the integral plant model for dynamic analyses of the WCLL BB configuration has been updated with special attention to primary system components. Following trends of relevant neutronics modelling, a new BB model has been implemented in 2020 with the aim to obtain higher resolution output data and a more realistic thermalhydraulic feedback from the primary system. Once-through steam generator user components have been built based on CAD models conceived by BOP partners. Transient analyses have been performed providing a better picture regarding the behaviour of main components, e.g., the BB and the OTSGs, whilst highlighting possible ways to optimise the control scheme of the plant.

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).

Development of a RELAP5/MOD3.3 Module for MHD Pressure Drop Analysis in Liquid Metals Loops: Verification and Validation

Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. Computational tools are employed to lead fusion-relevant physical analysis, but a dedicated MHD code able to simulate all the phenomena involved in a blanket is still not available and there is a dearth of systems code featuring MHD modelling capabilities. In this paper, models to predict both 2D and 3D MHD pressure drop, derived by experimental and numerical works, have been implemented in the thermal-hydraulic system code RELAP5/MOD3.3 (RELAP5). The verification and validation procedure of the MHD module involves the comparison of the results obtained by the code with those of direct numerical simulation tools and data obtained by experimental works. As relevant examples, RELAP5 is used to recreate the results obtained by the analysis of two test blanket modules: Lithium Lead Ceramic Breeder and Helium-Cooled Lithium Lead. The novel MHD subroutines are proven reliable in the prediction of the pressure drop for both simple and complex geometries related to LM circuits at high magnetic field intensity (error range ±10%).