Solid Breeder Research Articles

A case for gross electricity producing compact fusion pilot plants

Abstract A case for compact gross electricity producing pilot plant is presented.
The feasibility of such a plant with a moderate fusion power that is capable of delivering
gross electricity to the grid is investigated. The physics and engineering considerations
of such power plants are elucidated. We show that for a fusion power of about 300 MW
with fusion gain of 5, a moderate plasma β with improved confinement regime is
required to prevent excessive transport power loss. The sensitivity analysis indicates a
wide enough parameter range where, the fusion power and fusion gain can meet their
target values. The constraints arising from the shielding, magnets and maintenance
are discussed. The feasibility of steady-state gross electricity production of 160 MW
is discussed using a helium-cooled solid breeder blanket with an intermediate energy
storage system. It is argued that such a plant has all key technical elements of DEMO,
albeit at a smaller scale, thereby providing strong technical basis for DEMO.

Thermal Analysis of HCSB Blanket Concept for a Moderate Sized Tokamak Fusion Reactor

The conceptual design of a helium-cooled solid breeder blanket for a moderate sized tokamak fusion reactor is presented to explore the feasibility of heat extraction for power conversion. A compact reactor is considered with a major radius of 3 m, fusion power of 200 MW, and modular blankets. A configuration of radially stacked breeder units with a nested C-shaped helium flow scheme that enables nearly uniform high temperatures of the helium outlet is proposed. A detailed thermal analysis of the central module of the blanket is presented. A model is developed to carry out the one-dimensional thermal transient analysis of the breeder unit, which is benchmarked with ANSYS simulation. This can be used as a tool for rapid scan of the conceptual design for a wide range of parameters. The calculations show that a minimum pulse length of 2000s or above is required to achieve a steady-state blanket temperature for efficient heat extraction.

The diffusion and desorption of tritium on Li2TiO3 (-133) surface from first-principles calculations

Li2TiO3 is one of the most promising solid breeder materials for nuclear fusion reactor. Diffusion on the surface is a vital step of the tritium release process. In this work, the diffusion behavior of tritium atoms on the (-133) surface and the desorption of T2 molecules from the (-133) surface are systematically investigated. Possible adsorption sites, local diffusion pathways, and their corresponding activation energies have been identified using first-principles calculations and the climbing image nudged elastic band (CI-NEB) method. The diffusivity and effective activation energy for tritium diffusion on the (-133) surface have been determined using kinetic Monte Carlo (KMC) simulations. The activation energy for T2 desorption from the (-133) surface has also been investigated. The effective activation energy for tritium diffusion on the (-133) surface is found to be 1.06 eV, whereas the activation energy for T2 molecule desorption is 1.46 eV. The (-133) surface exhibits a higher activation energy for tritium diffusion but a lower activation energy for T2 desorption compared to the (001) surface.

Experimental study on the pressure drop of helium purge gas in particle crushing pebble beds

The solid breeder blanket is a critical component in fusion reactor, where helium purge gas flows through the solid breeder pebble bed to carry out the tritium generated during the fusion process. The flow pressure drop of helium purge gas within the breeder pebble bed is a significant parameter affecting the design of the tritium extraction system. Previous studies have indicated that the helium flow in the breeder pebble bed conforms to the theory of porous media flow. However, due to potential pebble breakage during the plasma operation, the pressure drop characteristics of the helium flow in breeder pebble bed may change as the void structure changes. The objective of this study is to measure the variation in flow pressure drop of the breeder pebble bed under different pebble crushing conditions. The flow pressure drops of intact beds (Alumina, diameter 1-1.2 mm) and four groups with different pebble breakage rates (3%, 5%, 7%, 9%) are measured using the pebble bed pressure drop test facility. The following results are obtained through experimental research: (1) The Ergun equation, Foumeny equation, and Reichelt equation can all reasonably match the experimental results of intact pebble beds; (2) The pressure drop across the pebble bed increases with the increase in pebble breakage rate, reaching approximately 1.6 times that of the intact bed at a 9% breakage rate; (3) A correlation for predicting the pressure drop of the broken pebble bed is established by introducing the pebble breakage rate (η) into the Ergun equation, which can be used to determine the pressure drop variation within a conservative range of breakage rates.

High pressure phase transition and its influence on structural, mechanical properties and elastic anisotropy of [formula omitted]: An ab-initio study

Lithium carbide (Li2C2), a promising tritium breeding material with potential applications in magnetic confinement-based fusion reactor blankets, has drawn significant attention due to its recent utilization in lithium-ion batteries. In the current study, we employed first-principles calculations with van der Waals (vdW) force correction to explore the structural and mechanical properties of Immm and Pnma phases of Li2C2 subjected to external hydrostatic pressure. It is demonstrated that inclusion of vdW effects accurately predicts Immm→Pnma phase transition at 16 GPa, aligning closely with experimental transition pressure. In contrast, PBE-GGA calculations underestimate the same by about 21%. Current paper is the first comprehensive study of pressure evolution of mechanical properties and strength determining parameters of Li2C2. Interestingly, this inherently brittle material exhibits ductile characteristics beyond 7.5 GPa. The paper also offers a quantitative comparison of its strength parameters with other prominent tritium breeding materials. Finally, the strong directional dependence due to the presence of C2 dimer is investigated by analyzing the pressure variation of elastic anisotropy. The study has its relevance in assessing structural integrity of Li2C2 during inadvertent buildup of tritium and helium that would generate stress in the solid breeder.

Lithium lead titanate (Li2PbxTi1-xO3, 0.1

Tritium breeder is an important functional material in d-T fusion reactor blanket serving the purpose of tritium re-production and energy conversion. Presently, Li2TiO3 ceramics is considered as the most promising solid tritium breeder for its good chemical stability, high mechanical strength and excellent tritium release performance. However, low lithium content and lithium evaporation at high temperature lead to a low tritium breeding ratio. In the paper the neutron multiplier element, Pb, is introduced into Li2TiO3 to obtain lithium lead titanate (Li2PbxTi1-xO3) as tritium breeder, which combined the advantages of Pb and Li2TiO3 ceramic. Li2PbxTi1-xO3 powders were synthesized via a sol-gel method, employing lithium acetate, tetrabutyl titanate, and lead acetate, with varying lead contents. The characterization results indicate that at a molar ratio of titanium to lead of 1:2, Li2TiO3-Li2PbO3 eutectic phase could be formed. The tritium breeding ratio (TBR) of Li2PbxTi1-xO3 materials were evaluated by Monte Carlo Code, which was notably influenced by the lead content, exhibiting a continuous increase with higher Pb content and neutron energy. The Li2PbxTi1-xO3 pebbles were prepared by freeze-drying route. The mechanical strength of Li2PbxTi1-xO3 pebbles could be affected significantly by the lead content. Compared to the conventional Li2TiO3 ceramics, the Li2Pb0.4Ti0.6O3 ceramics exhibits a higher crushing load of 25.33 N.

The effect of Li2TiO3 pebbles on the mechanical properties of ARAA under breeding blanket conditions

The breeding blanket serves as a critical element for achieving a self-sustaining tritium cycle and generating sufficient energy in fusion reactors. This study focuses on candidate materials of the helium-cooled solid breeder blanket designed for a demonstration fusion reactor breeding blanket in Korea. We conducted compatibility tests between advanced reduced-activation alloy (ARAA) and Li2TiO3 pebbles at 550 °C for 28 days in He-0.1 vol% H2 purge gas conditions. Surface oxidation on ARAA was observed regardless of pebble presence. Double oxide films comprised of Fe-rich and Cr-rich layers formed on the surface of ARAA with pebbles. Lithium deposition was verified in both contact and non-contact areas of ARAA. Despite surface oxidation, the impact on tensile strength and elongation was found to be negligible.

Macroscopic transport characteristics in packed bed with a multi-physics inversion strategy for solid breeding blanket of HCCB TBM

The solid breeding blanket with typical packed bed structure is a key component for tritium breeding and thermal conversion in future fusion reactors. The pore scale model for multi-physical field solution of solid breeding blanket is fine but very expensive. The simplified packed-bed scale model is more favored in engineering especially for the system-level calculation of the blanket, but it needs to be inputted appropriate macroscopic transport parameters. In this work, the multi-physics calculations of blanket in Chinese HCCB TBM (Helium-Cooled Ceramic-Breeder Test Blanket Module) at pore scale and packed-bed scale were performed. It was found that the packed-bed scale model with traditional macroscopic transport parameters is not accurate enough, especially the prediction deviation of tritium concentration distribution could reach about 25 %. Hence, a novel multi-physics inversion strategy was proposed to predict the macroscopic transport parameters of the packed bed under the multi-field coupling conditions. The inversion results show that the convection of helium significantly enhances the heat and tritium diffusion in the packed bed, and the effective thermal conductivity and effective diffusivity could be increased by 18 % and 46 %, respectively. Then based on the results by multi-physics inversion strategy, the modified effective thermal conductivity and effective diffusivity correlation models for solid breeding blanket packed bed are constructed. With the modified correlation models, the relative prediction deviations of pressure drop, temperature distribution and concentration distribution are within 2.5 %, 0.6 % and 10 % respectively at different inlet velocities (0.05 m/s ∼ 0.25 m/s) and different inlet temperatures (300 K ∼ 900 K). The study in this paper would assist the thermal-hydraulics analysis of the solid breeding blanket in the fusion reactor.

Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall

The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various fabrication methods have been studied for the development of the FW fabrication technologies. This study focuses on FW fabrication technologies with hot isostatic pressing (HIP) bonding which is considered as the most promising method for achieving such complex shape like FW having internal cooling channels. We investigated the HIP bonding conditions using Advance Reduced Activation Alloy, ARAA, which is a reduced activation ferritic-martensitic steel under development in Korea. Temperature, holding time and pressure were considered as the parameters to optimize the HIP bonding conditions. The microstructures around the HIP joints changed with respect to the HIP bonding temperature. The amount of oxide particles around the HIP joints varied depending on the degassing conditions. The tensile strength of the HIP joint was not greatly affected by the presence of the HIP bonding line and degassing conditions. On the other hand, it has been confirmed that the Charpy absorbed energy of the HIP joint was significantly affected by degassing conditions. An increase in oxide particles around the HIP joints occurred when degassing temperature and vacuum level were low, and as the quantity of oxide particles increased, the Charpy absorbed energy of the HIP joint decreased substantially.

Molecular dynamics simulations of the effect of porosity on heat transfer in Li2TiO3

Heat transfer is a key consideration in the development of tritium breeder blankets for future fusion reactors. For solid tritium breeder materials there is a fine balance to be struck between high levels of porosity to encourage tritium release and minimising it to maintain the thermal and mechanical properties. Therefore, in this work we employ molecular dynamics simulations to understand how the introduction of porosity influences the thermal conductivity of lithium metatitanate ceramic breeder material. Our simulations predict that increasing the porosity leads to a decrease in the thermal conductivity which is in good agreement with previous experimental observations. By contrast, we do not observe the increase in the thermal conductivity at high temperatures, that is observed in some experiments. We argue that this increase is a consequence of sintering or some other modification of the experimental sample rather than a fundamental change in the heat conduction mechanism in the crystal matrix.

Design and Development of Hydrogen Isotopes Extraction System at IPR

Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles Li2TiO3 (Lithium Titanate) as solid breeders and second is the lithium based alloy Pb-Li (Lead Lithium) as a liquid breeder. Tritium bred in solid breeders is extracted with purge helium gas, whereas tritium bred in liquid Pb-Li needs suitable gas-liquid contactor. It should be noted that, tritium recovery (i.e.,-extraction) from Pb-Li is very challenging due to low solubility of tritium in liquid Pb-Li and no such separation technology readily available commercially. Therefore, the necessity for developing a reliable and efficient tritium recovery technology (i.e.,- extractor) for liquid Pb-Li is of key interest.In present work, design and development of Hydrogen Isotopes Extraction System (HIES) for hydrogen recovery from liquid Pb-Li is discussed. As hydrogen and tritium are isotopes, hydrogen is used instead of tritium in present system due to several safety concerns with tritium. Recovery of hydrogen from liquid Pb-Li is carried out using structured packing as a Gas-Liquid Contactor (GLC). The design details of gas-liquid contactor (i.e.-extractor column) and various process tanks of HIES are discussed. In present design, configuration of vertically mounted extractor column makes overall HIES relatively tall (∼ 3 m height) and use of liquid Pb-Li as process fluid makes the system heavy weight (∼1500 kg) too. In addition to this, system needs to be operated above Pb-Li melting temperature (> 235 C) to keep Pb-Li in liquid form. Therefore, thermal analysis of piping as well as seismic analysis of whole system is carried out and discussed along with various operational stages of HIES.

Experimental Investigation on Pressure Drops of Purge Gas Helium in Packed Pebble Beds for Nuclear Fusion Blanket

The flow characteristics of purge gas helium in the pebble bed of the tritium breeding blanket are important in analyzing the tritium purging process and optimizing the design of the solid breeder blanket. Therefore, the flow characteristics of helium gas in randomly packed pebble beds are investigated experimentally with a focus on the analysis of the pressure loss distribution. The results show that gas velocity, bed dimension, and pebble diameters have an obvious influence on the helium flow characteristics in pebble beds. With the increase in the inlet helium gas velocity, the pressure drop gradient of helium in the pebble bed gradually increased. With increases in the pebble bed dimension, the pressure drop gradient of helium in the pebble bed gradually increased. With the increase in the pebble diameter, the pressure drop gradient gradually decreased. In addition, the effect of temperature on the pressure drop of helium in the pebble bed was also preliminarily investigated. The pressure drop gradient of the helium through the pebble bed obviously increased with the increase in the helium and the bed temperature. The experimentally obtained pressure loss characteristics can be used for the validation of the simulation of a blanket pebble bed and as input parameters in the thermo-hydraulic analysis of solid-tritium breeder blankets.

Development of water-cooled cylindrical blanket in JA DEMO

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

High-temperature interaction of water vapor with lithium ceramics Li2TiO3

Selection of materials for fusion reactors and, in particular, for blanket systems is one of the most important tasks that determine the development of fusion energy. Lithium ceramics Li2TiO3 is considered as prospective candidate for solid breeders of future fusion reactors' blankets. The paper presents the results of high-temperature tests of Li2TiO3 + 5 mol% TiO2 ceramics, performed under conditions of blowing with argon containing impurities of water vapor of different isotopic compositions (H2O, HDO, D2O) in the range of temperature from 100 °C to 1400 °C. The vapor–gas mixture for the experiment was prepared in a special mixing tank. The interaction of Li2TiO3 + 5 mol% TiO2 ceramics with chemically active gases and water vapors was investigated by thermogravimetry (TG), differential scanning calorimetry (DSC) and mass spectrometry (MS) methods. As a result of experiment, the dependences of the sample mass change during heating under single purge gas composition was obtained. Mass spectra that characterize the changes in the qualitative composition of the gas mixture in the reaction chamber with the investigated sample were recorded and subsequently analyzed. An isotope effect is observed during the experiment: the release of D2 gas begins significantly earlier than the release of HD gas. The observed effect appears only in the high-temperature region (above 1050 °C), which characterizes the investigated material as almost completely chemically neutral towards water vapor.

Solid breeder material crystal structure evolution due to Li burn up–Loss of crystal stability

Solid breeder materials for fusion power plants are subjected to extreme conditions of radiation damage, high temperature and designed component burn-up. The loss of lattice stability of Li2TiO3 due to Li burn-up was investigated using first principles-based approaches to aid evaluation of upper limit of component lifetime. The lattice stability is analyzed using calculated phonon dispersion in the Li2TiO3 supercell, while Li burn-up is modeled by the introduction of Li-vacancies. It has been determined that the studied ceramic can be structurally stable up to ∼40% of Li atoms burn-up. At 50% burn-up, the negative frequency branch in the phonon dispersion spectrum appears indicating the loss of lattice dynamical stability. Moreover, the structure obtained by introduction of unstable frozen phonons with minima energy amplitude also results in a phonon dispersion with negative branches in the structure obtained. Therefore, the system completely loses stability when approximately half of Li atoms are consumed.

Investigation of the size-dependent crushing behavior of Li4SiO4 pebbles via discrete element method

Li4SiO4 pebbles are commonly employed as breeder materials in solid breeder blankets. As a typical brittle material, the Li4SiO4 pebble sizes significantly affect their crushing behavior. In this study, the size-dependent crushing behavior of pebbles is investigated via discrete element method. Based on the Weibull theory, the probabilistic crushing strength of Li4SiO4 pebbles is analyzed. The numerical results demonstrate that both the characteristic crushing strength and Weibull modulus display a negative linear relationship with Li4SiO4 pebble size in a double-logarithmic plot. Furthermore, by considering both the axial force-displacement responses and fracture process of pebbles, three typical failure patterns of Li4SiO4 pebbles can be identified, namely, splitting failure (Pattern-I), progressive fracturing failure (Pattern-II) and explosive failure (Pattern-III). As the pebble size decreases, the failure pattern gradually shifts from Pattern-I to Pattern-II and to Pattern-III. Additionally, the fragment size distributions of crushed Li4SiO4 pebbles are analyzed based on the fractal theory. It is observed that the fragment size distributions of Li4SiO4 pebbles with varying sizes exhibit favorable fractal features. Smaller Li4SiO4 pebbles exhibit a more significant fragmentation, as evidenced by the increasing fractal dimension of the fragment size distribution with the decreasing pebble size. This work may provide valuable support for further investigations into the mechanical performance of a Li4SiO4 breeder blanket using DEM simulations.

Enhancing the density and crush load of Li2TiO3 tritium breeding ceramic pebbles by adding LiNO3-Li2CO3

The fabrication of Li2TiO3 ceramic pebbles with satisfactory density and excellent crush load while maintaining a fine grain structure is one of the research objectives of solid tritium breeders. However, densification and grain refinement are often contradictory in lithium-containing ceramics during the sintering process. In this work, LiNO3-Li2CO3 was added during the preparation of Li2TiO3 green spheres by the wet process, and the formation of LiNO3-Li2CO3 molten salt resulted in the significant sintering improvement of Li2TiO3 ceramic pebbles. The addition of LiNO3-Li2CO3 increased the density of Li2TiO3 ceramic pebbles by 10.47 % (92.63 % vs. 82.16 %) and the crush load by more than four times (144.54 N vs. 28.2 N). ICP-OES analysis showed that the lithium content of the Li2TiO3 ceramic pebbles was slightly increased by the addition of LiNO3-Li2CO3. Overall, the difficulty of balancing grain refinement and densification in lithium-containing tritium breeders can be addressed by the proposed method.

Reactor experiments on irradiation of two-phase lithium ceramics Li2TiO3/Li4SiO4 of various ratios

The design of future fusion reactors involves the production of tritium inside the breeder blanket. The most promising material for solid breeder blankets is a two-phase lithium ceramic containing orthosilicate Li4SiO4 (LOS) and metatitanate Li2TiO3 (LMT) of lithium in various proportions. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. Further, this tritium is extracted from the blanket with a purge gas and returned to the fusion zone, realizing the concept of a closed fusion cycle.Irradiation under fission reactor conditions is still one of the few available methods for estimating the parameters of tritium generation and release from lithium-containing materials in the "in-situ" mode. This paper presents the results of experiments on neutron irradiation of two-phase lithium ceramics of various ratios (LOS + 35 mol. % LMT (pebble size 250–1250 μm), LOS + 35 mol. % LMT (pebble size 500–710 μm) and LOS + 25 mol. % LMT (pebble size 500–710 μm) at the WWR-K research reactor.Irradiation of each batch of samples lasted from 5 to 22 days. The experiments were carried out by the vacuum extraction method. This paper describes the main methodological aspects of the studies, namely the technical features of four irradiation campaigns, the sequence and scope of the studies. A comparison is also made of the initial sections of reactor experiments for all campaigns, where the reactor was sequentially brought to power, according to which the parameters of the Arrhenius dependence of the effective tritium diffusion coefficients were estimated.

Features of Helium and Tritium Release from Li2TiO3 Ceramic Pebbles under Neutron Irradiation.

The operation of fusion reactors is based on the reaction that occurs when two heavy hydrogen isotopes, deuterium and tritium, combine to form helium and a neutron with an energy of 14.1 MeV D + T → He + n. For this reaction to occur, it is necessary to produce tritium in the facility itself, as tritium is not common in nature. The generation of tritium in the facility is a key function of the breeder blanket. During the operation of a D-T fusion reactor, high-energy tritium is generated as a result of the 6Li(n,α)T reaction in a lithium-containing ceramic material in the breeder blanket. Lithium metatitanate Li2TiO3 is proposed as one of the promising materials for use in the solid breeder blanket of the DEMO reactor. Several concepts for test blanket modules based on lithium ceramics are being developed for testing at the ITER reactor. Lithium metatitanate Li2TiO3 has good tritium release parameters, as well as good thermal and thermomechanical characteristics. The most important property of lithium ceramics Li2TiO3 is its ability to withstand exposure to long-term high-energy radiation at high temperatures and across large temperature gradients. Its inherent thermal stability and chemical inertness are significant advantages in terms of safety concerns. This study was a continuation of research regarding tritium and helium release from lithium metatitanate Li2TiO3 with 96% 6Li during irradiation at the WWR-K research reactor using the vacuum extraction method. As a result of the analysis of experiments regarding the irradiation of lithium metatitanate in vacuum conditions, it has been established that, during irradiation, peak releases of helium from closed pores of the ceramics are observed, which open during the first 7 days of irradiation. The authors assumed that the reasons samples crack are temperature gradients over the ceramic sample, resulting from the internal heating of pebbles under the conditions of their vacuum evacuation, and contact with the bottom of the evacuated capsule. The temperature dependence of the effective diffusion coefficient of tritium in ceramics at the end of irradiation and the parameters of helium effusion were also determined.

In-situ tritium release behavior and post-irradiation experiments of Li2TiO3 pebbles developed by freeze-drying method

To assess the overall performance of potential solid tritium breeders for China Fusion Engineering Test Reactor (CFETR), we conducted a series of In-situ irradiation experiments on the in-pile tritium production and extraction experimental platform at China Mianyang Research Reactor (CMRR). In 2019, we performed an in-pile tritium release experiment (IPTRE-2) on Li2TiO3 (∼90 T.D%, 6Li 7.5%, ∼30.17 g) ceramic pebbles, which are the latest developed breeders. The Li2TiO3 pebbles were produced at the Institute of Nuclear Physics and Chemistry (INPC), China Academy of Engineering Physics (CAEP), using a freeze-drying method. This paper reports on the In-situ tritium release behavior and the post-irradiation experiments of Li2TiO3 pebbles. We established the relationship between tritium residence time and temperature, which can be expressed as lgτ=-0.74+1622.4/T. Our findings indicate that the Li2TiO3 pebbles remained structurally stable without fragmentation, but their color darkened and the degradation of mechanical properties due to the reduction of Ti in Li2TiO3.