Solid Breeder Blanket Research Articles

High lithium content silicates: A comparative study between four routes of synthesis

Lithium-silicate ceramic is a material widely used in the energy field, here studied for its application as solid breeder blanket in the future fusion reactors. Considering the difficulty in reaching a unique stable phase together with a spherical shape, the advantages and disadvantages of solid-state reaction, drying gels in rotary evaporator, spray drying and reflux in alcoholic media methods are here considered for the production of lithium orthosilicate (Li4SiO4). The fabrication of a material with high Li-content such as the lithium oxosilicate (Li8SiO6) is also investigated.Phase compound identification, mass and heat evolution are followed by X-Ray Diffraction, thermo-gravimetrical analysis (DTA/TG) and infrared spectroscopy (ATR/IR) while the samples morphology is observed by the Scanning Electron Microscope.The spray drying method results to be the best fabrication way to achieve the desired compromise between shape and crystallinity in the case of the orthosilicate ceramic. The oxosilicate phase is reached by solid-state route and in small amount, by spray drier and reflux processes.

A Conceptual Study on Helium-Cooled Solid Breeder Blanket for Prototype CH DEMO

A conceptual design of tritium-breeder blanket for prototype DEMO is introduced. In this concept, Li4SiO4 pebbles are chosen as tritium-breeder material and Be for neutron multiplication. To confirm the thermal effectiveness,a CFD code FLUENT is used for the calculation. The result indicated that, the thermal effectiveness of the concept is acceptable.

Recent developments of solid breeder fabrication

Worldwide research and development has been performed on solid breeder materials as a tritium source for future fusion power reactors. These ceramic materials will be used in the International Thermonuclear Experimental Reactor (ITER) to check their performance and screen their applicability for use in fusion blankets. While all members of the ITER project are engaged in the development of solid breeder blanket concepts, five are actively involved in the fabrication of tritium breeding ceramics for the different test blanket modules to be tested in ITER.All solid breeder concepts proposed for ITER are based on the use of ceramics in the form of pebble beds with pebble diameters of about 1mm or smaller. Although alternative materials are still under investigation, the main candidates for tritium breeding ceramics are lithium metatitanate and lithium orthosilicate. Recent developments and preparatory work for ITER are presented.

Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept

India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li2TiO3/Li4SiO4 and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal–hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature–time histories. Result of thermal–hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis will be presented in this paper.

Preliminary Evaluation of the Adequacy of Lithium Resources of the World and China for D-T Fusion Reactors

This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the “permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

Status of ceramic breeder pebble bed thermo-mechanics R&D and impact on breeder material mechanical strength

Among the international fusion solid breeder blanket community, there exists steady progress on the experimental, phenomenological, and numerical characterizations of the pebble bed effective thermo physical and mechanical properties, and of thermomechanic state of the bed under prototypical operating conditions. This paper summarizes recent achievements in pebble bed thermomechanics that were carried out by members of the IEA Fusion Nuclear Technology Subtask I Solid Breeding Blanket. A major goal is on developing predictive capability while identifying a pre-conditioned equilibrium stress state that would warrant pebble bed integrity during operations. The paper reviews and synthesizes existing computational modeling approaches for pebble bed thermomechanics prediction, and differentiating points of convergence/divergence among existing approaches. The progress toward modeling benchmark is also discussed. These advancements have led to a framework to help navigate future research.

Development of a technical scale PERMCAT reactor for processing of highly tritiated water

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·1016Bqkg−1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D2O and H2, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.

Fabrication and characterization of Li4SiO4 pebbles by melt spraying method

Abstract The solid breeder blanket concept proposed by the China features the tritium breeding ceramic as pebble beds in several submodules. The lithium orthosilicate (Li4SiO4) is considered as first candidate ceramic breeder materials fabricated by the melt spraying method, which is favorable to other processes in terms of density and recycling. The production process involves rapid quenching of the liquid droplets from the melt to room temperature which cause internal stresses and leads in some cases to formation of microcracks and the dispersion of mechanical properties. Molar ratio (Li/Si) of the pebbles was evaluated by ICP–OES. It is shown that the Li/Si ratio of the pebbles is slightly varying from batch to batch because of evaporation of lithium at high temperatures. The crush tests on single pebbles show that a mean value of 7.0 N was obtained in crush load experiments of 40 pebbles with a diameter of 1.0 mm. It results that heat treatment of pebbles improves the density and mechanical stability. The activation characteristics for the current composition of Li4SiO4 pebbles were assessed. The calculations were used to identify critical amounts of impurities and were compared to the results of pure material without impurities.

Preliminary neutronic assessment of a helium-cooled Li8PbO6 breeding blanket design for DEMO

A preliminary neutronic assessment of the performances of a helium-cooled Li8PbO6 breeding blanket for the conceptual design of a DEMO fusion reactor is given. The study mainly focuses on TBR, power density responses and shielding factor optimization to estimate the feasibility of the design under the prescribed radiation deposition limits at TF-coils superconducting magnets. Computational analyses are based on three-dimensional 30° sector using the Monte Carlo code MCNPX 2.6. The scoping interest of helium-cooled Li8PbO6 blanket designs is based on a large potential minimization of the amount of Be required and the strong relaxation of 6Li enrichment requirements for this solution when compared to other solid breeder blanket options.

Enhanced fabrication process for lithium orthosilicate pebbles as breeding material

The solid breeder blanket concept proposed by the European Union features the tritium breeding ceramics as pebble beds inside several breeder units. The pebble bed will not take part in the structural rigidity of a breeder unit. Still the pebbles need to withstand forces originating from thermal expansion mismatches and neutron irradiation. A designated breeder ceramic for this concept is lithium orthosilicate with a slight surplus of silica. It has been demonstrated that lithium orthosilicate can be processed by melt spraying, which is favorable to other processes in terms of recycling. This work summarizes changes to the process and investigations of the current limitations that have to be overcome to enhance not only the mechanical properties of the pebbles, but also the yield of the process. A key feature of the improvements is to control the forming of the single pebbles more precisely than in the past. Thereby, not only the generation of pores should be avoided, but also the size distribution of the pebbles should be more accurately controlled.

IFMIF suitability for evaluation of fusion functional materials

The International Fusion Materials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusion materials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design.This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functional materials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor.

Blanket Concept of Water-Cooled Lithium Lead with Beryllium for the SlimCS Fusion DEMO Reactor

As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket.

DT中性子照射による固体増殖ブランケット模擬体系からのトリチウ ム回収実験 -(1)

DT中性子照射による固体増殖ブランケット模擬体系からのトリチウ ム回収実験 -(1)

Crush probability analysis of ceramic breeder pebble beds under mechanical stresses

A framework for analyzing crush events of individual ceramic pebbles in solid breeder blankets is developed by means of probabilistic methods. As a brittle material, ceramic breeder pebbles show considerable scatter in crush strengthen for single pebbles. The combination of the discrete element method and experimental data of crush loads provides the possibility of obtaining the overall crush probability of a pebble bed under compression. Furthermore, micro–macro relations are used to correlate the crush probability of pebbles with the overall stress level of the bed. Analysis of uniaxial compression of a mono-sized lithium-orthosilicate pebble bed is presented to demonstrate the application of this tool.

Theoretical calculation and analysis modeling for the effective thermal conductivity of Li 4SiO 4 pebble bed

The effective thermal conductivity of Li 4SiO 4 pebble bed is an important design parameter and must be known for the thermo-mechanical design of solid breeder blankets. The theoretical calculation and modeling analysis for the effective thermal conductivity of Li 4SiO 4 pebble bed are performed in this paper. The 2D and 3D theoretical equations for the thermal conductivity of Li 4SiO 4 pebble bed are given, moreover compared with the modeling results and other experimental data. The results show that the effective thermal conductivity of Li 4SiO 4 pebble bed can be preliminarily obtained by analysis modeling or theoretical calculation under the lack of experimental set-up at present.

Integrated simulation of tritium permeation in solid breeder blankets

Numerical simulation of co-permeation of tritium and hydrogen from breeding zones to the coolant in the helium cooled pebble-bed blanket is performed in this paper. 3D multi-species convection–diffusion models integrated with thermal-fluid analysis in porous media are assessed and then used to estimate the associated tritium permeation for a solid breeder blanket module. Benchmark calculations give a reasonable agreement on the co-permeation rates with the experimental data. Simulation in a TBM unit show that purge gas flow can strongly affect tritium transport, increasing the purge flow velocity is an effective method to reduce tritium permeation to the coolant. In the case where hydrogen is added to the purge gas stream to promote tritium release, the co-permeation of H 2, T 2, and HT are taken into account in the permeation simulation, results show that permeation flux of T–T molecules is reduced due to the effect of co-permeation of hydrogen.

Search for reality of solid breeder blanket for DEMO

For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li 4SiO 4, Li 2TiO 3 and Li 2ZrO 3) and Be multipliers (Be and Be 12Ti). Based on a neutronics-heat coupled analysis, the tritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900 °C for Li ceramics and Be 12Ti, and ≤600 °C for Be. Cooling water condition was assumed to be 23 MPa and 290–360 °C. The result indicates that surplus tritium production in lower neutron wall load ( P n ) blanket compensates a shortfall in higher P n blanket and thus the overall tritium production can marginally satisfy fuel self-sufficiency.

Comparison of Two Types of Blanket Concepts for a Fast Ignition ICF Reactor FALCON-D

A Fast ignition Advanced Laser fusion reactor CONcept with a Dry first-wall and a high repetition laser (FALCON-D) has been proposed to investigate the potential of the fast ignitionin the reactor concepts. For the blanket system, two types of blanket concepts, i.e.asolid and a liquid metal breeder types using the reduced activation ferritic steel (F82H) were proposed. In this study, two types of blankets were designed, where the thickness of the blankets was minimized while keeping the net TBR larger than 1.07. One of the blanket concepts for FALCON-D is based on the solid breeder (Li2TiO3) with beryllium (Be) neutron multiplier and water cooling. The second blanket concept is based on liquid metal breeder (Li17Pb83) with water cooling. The maintenance method for FALCON-D is applicable to both blanket types. The net electric power of the solid breeder blanket is 110 MW larger than that of the liquid metal breeder blanket. This is mainly caused by the differences in the neutron energy multiplication. In the case of the liquid metal breeder blanket with water cooling, the net TBR 1.09 is achieved without Be as the neutron multiplier. Such design without Be can remove a risk of accident due to the chemical reaction between beryllium and water. From the economical point of view, the solid breeder blanket with water cooling, which generates a larger electric power, is desirable. On the other hand, if the combination of beryllium and water cooling was not acceptable from a viewpoint of safety, the blanket system with the liquid metal would be another possible option.

Current R&D Activities on Korean Helium Cooled Solid Breeder Test Blanket Module

Helium Cooled Solid Breeder (HCSB) blanket is one of two blanket concepts being considered as Korean Test Blanket Module (TBM) for ITER with the aim for testing and verifying the capability of the breeding blanket concepts. R&D activities being carrying out for HCSB TBM include the development of materials, fabrication technologies, and TBM associated system design. A small sample of ferritic/martensitic (FM) steel was fabricated. It was found that the tensile strength was close to previous value. A fabrication technique of sphere pebbles of lithium titanate breeder and graphite reflector was developed. A FM/FM TIG welding was performed and the results showed that tensile strength of the welded zone was decreased about 10 %. A small punch test method of mechanical property evaluation was introduced to verify the suitability of small specimen for irradiation test by examining the relationship of the conventional uniaxial tensile test, and the tensile strengths were compared. As TBM design has complicated square channel configuration, short square channel was fabricated successfully. Finally, the components and specifications of the TBM associated systems are described in this paper.

Numerical Analysis for Thermal and Mechanical Properties of Pebble Bed

In solid breeder blanket design of fusion power plants, the ceramic breeder pebble bed plays a very important role. Its mechanical and thermal properties are necessary to design the blanket. In this study, thermomechancal properties of the bed such as effective thermal conductivity and stress-strain relation are investigated by means of numerical simulation. A discrete element manner is adopted in the simulation in order to clarify the influence of the individual particle properties upon the bulk behavior. As a result, the thermo-mechanical properties of pebble bed were well re-created in a computational space by means of the numerical model used in this study.