Ceramic Breeder Research Articles

Tritium release behavior of Li2TiO3 and 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles fabricated by microwave sintering

Li2TiO3-Li4SiO4 biphasic ceramic is considered as an advanced tritium breeder material for its potential to combine the advantages of both Li2TiO3 and Li4SiO4. In this study, Li2TiO3 and 2Li2TiO3-Li4SiO4 biphasic pebbles with promising properties were fabricated by microwave sintering. The 2Li2TiO3-Li4SiO4 biphasic pebbles displayed porous structure with a grain size of 170 nm. Tritium release behavior for the microwave sintered ceramic breeder pebbles was investigated. It was found that the microwave sintered Li2TiO3 pebbles show the same release behavior with that of conventional sintered pebbles due to the similar microstructure and density properties. Besides, the tritium release results indicated that high closed porosity may lead to the increase of release temperature. The 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles hold satisfactory tritium release behavior with a low release temperature around 450 °C. According to the tritium release results, it could be concluded that the interfaces between Li2TiO3 and Li4SiO4 can improve the tritium release behavior by promoting the surface reaction.

Microstructure and mechanical properties of LBW and EBW weld joints of CLF-1 steel: A comparative analysis

This study was focused on welding joints of reduced-activation ferritic/martensitic (RAFM) CLF-1 steel medium-to-thick plates produced via laser beam welding (LBW) and electron beam welding (EBW). Such joints were used in the ITER project, in particular, in the back plate mockup of Helium Cooled Ceramic Breeder Test Blanket System developed in China in 2015. The microstructural evolutions and the mechanical properties of fusion zones (FZs), weld zones (WZs), and heat-affected zones (HAZs) of LBW and EBW joints were analyzed and compared using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Moreover, the microstructural differences between the two types of joints and their effect on the impact properties were discussed. Results showed that the well-formed welds without defects, such as pores, incomplete fusion, and cracks, could be obtained via LBW and EBW. Different heat inputs led to the differences in the grain size gradients, the boundary angles, the content of carbides, and the distribution of dislocations. Despite obvious differences in morphology among FZ, WZ, and HAZ, both types of joints exhibited excellent tensile properties and impact toughness. The tensile strengths of the LBW and EBW joints were 633 and 634 MPa at room temperature, respectively, versus 370 and 366 MPa at 550 °C. The mean impact-absorbing energy values of LBW and EBW joints were 251 and 284 J. The further analysis revealed that the low δ-ferrite content, small effective grain size, large grain size gradient, high grain boundary angle, diminutive size of M23C6 carbides, high-Ta content in MX carbides and low dislocation density led to high impact toughness of the LBW joint, while the microstructure of EBW joint failed to show the improvement of impact toughness, and the extraordinary impact toughness could be attributed to the high scattering band.

HCCR breeding blankets optimization by changing neutronic constrictions

The neutronic analysis of Helium Cooled Ceramic Reflector (HCCR) breeding blankets has been performed using the 3D Monte Carlo code MCNPX and ENDF nuclear data library. This study aims to reduce 6Li percentage in the breeder zones as much as possible ensuring tritium self-sufficiency. This work is devoted to investigating the effect of 6Li percentage on the HCCR breeding blanket's neutronic parameters, such as neutron flux and spectrum, Tritium Breeding Ratio (TBR), nuclear power density, and energy multiplication factor. In the ceramic breeders at the saturated thickness, increasing the enrichment of 6Li reduces its share in the tritium production. Therefore, ceramic breeders typically use lower enriched Li from 30% to 60%. The investigation of neutronic analysis in the suggested geometry shows that using 60% 6Li in Li2TiO3 can yield acceptable TBR and energy deposition results, which would be economically feasible.

Comparative activation analyses for the HCPB breeding blanket in DEMO

The Helium Cooled Pebble Bed (HCPB) breeding blanket is one out of two concepts considered as driver blanket for the European DEMOnstration fusion power plant. The latest design, developed within the Breeding Blanket project of EUROfusion’s Power Plant Physics and Technology (PPPT) programme, assumes a fission-like breeder pin arrangement with breeder ceramics and Be12Ti as neutron multiplier.For the assessment of safety and waste management issues related to the use of the HCPB blanket in DEMO, a thorough activation analysis is required. Such an analysis should provide the total activity inventory and decay heat as well as the breakdown into components and materials as resulting from the irradiation over the assumed blanket lifetime in DEMO. An accurate calculation of the activity inventories requires the application of a suitable computational approach coupling the neutron transport simulation and the activation calculation with fully heterogeneous 3D geometry model of the DEMO.Precise materials specifications, including impurities and tramp materials, are essential prerequisites for adequate activation calculations. For the materials used in the HCPB blanket, the material compositions are provided with the material specifications set up for PPPT neutronics and activation calculations. These include the Eurofer steel, the breeder ceramics, tungsten and the beryllium in the Be12Ti used as neutron multiplier.Two difference tritium breeding material compositions of activation analyses were performed using each Li4SiO4+Li2TiO3 the latest material and previous material definition Li4SiO4. And beryllium titanate, the activation calculations were performed using the impurities specifications as provided by different potential beryllium suppliers. The most essential here is a uranium content in the ore used for the Be12Ti metallurgical engineering. This can permit a justification of the potential environmental risks associated with the application of the various materials options. The effect of the different Lithium compositions on the activity, both on the short- and long-term time scale, is evaluated in this work in the context of the radioactive waste management issue for DEMO.

Preliminary thermal-hydraulic influence of the poloidal nonuniform neutron wall loading on the CFETR HCCB blanket

Abstract Helium Cooled Ceramic Breeder (HCCB) blanket, which has been determined as one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR), is the critical component of the fusion reactor. The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation, and they are also the initial boundary conditions for the other performance analyses. Both performances are influenced by the neutron wall loading (NWL) distribution on the blanket and the radial structural arrangement of the internal functional zones. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the blanket have been investigated preliminarily. However, as the nuclear power deposition in each blanket component is the internal heat source for the thermal-hydraulic analyses, it’s essential to further investigate the influence of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. In this paper, taking the typical CFETR HCCB blanket module as the research object, firstly, both the neutronics and the thermal-hydraulic models are established. For investigating the influence of the poloidal nonuniform NWL on blanket performances, both models are divided into five identical parts along the poloidal direction. Afterwards, by referencing the actual poloidal NWL distribution in CFETR, we have aggregately proposed three kinds of poloidal nonuniform NWL distribution forms, including two symmetrical and one stepped NWL. On these bases, both the neutronics and the thermal-hydraulic analyses under the poloidal nonuniform NWL distributions are conducted, and the calculation results are compared with those corresponding to the uniform NWL distribution. The comparison results show that the maximum temperature of every functional zone calculated under the three nonuniform NWL distributions is generally higher than that corresponding to the uniform NWL distribution, which means that the thermal-hydraulic results calculated under the simplified uniform NWL distribution are not conservative. Under the condition of nonuniform NWL distributions, there is obvious poloidal temperature nonuniformity in every functional zone. Besides, the temperature deviations between the stepped and the uniform NWL distributions are generally larger than those between the symmetrical and the uniform ones. This paper can provide some valuable references for the detailed engineering design, analyses, and optimizations of the CFETR HCCB blanket.

Chemical compatibility between the α-Al2O3 tritium permeation barrier and Li4SiO4 tritium breeder

Abstract α-Al2O3 tritium permeation barriers were exposed in a 1-MPa He+0.5-MPa H2 atmosphere to Li4SiO4 ceramic tritium breeder at 623 K for 120–1000 h, to evaluate its chemical compatibility in a simulated breeder blanket environment. The as-deposited coating is composed of the α-Al2O3 phase with a uniform and dense surface morphology. The short-term (below 360 h) effects of corrosion are almost negligible. However, the morphology and structure of the coating change significantly when the corrosion time reaches 1000 h, and grain boundary corrosion, delamination, holes and corrosion products of LiAlO2 are observed. A microscopic model is proposed to understand the corrosion mechanism of the α-Al2O3 coating in contact with the Li4SiO4 ceramic.

Compatibility of tritium permeation barrier coatings with ceramic breeder pebbles

Reduced activation ferritic/martensitic steel F82H samples with tritium permeation barriers were annealed in contact with ceramic breeder pebbles at 550 °C and 700 °C for up to 32 days under purge gas. Cr2O3-formed F82H rapidly reacted to form three layers at 550 °C, and pure Fe projections grew with annealing time. Diffusion kinetics of O and Li were well fitted with Fick’s first law of diffusion. Er2O3 coatings did not react with the pebbles, but deteriorated by oxygen diffusion through the coatings. The impurity concentrations in the system and the coating thickness affect the coating compatibility in solid breeder blanket environments.

Current status and future perspectives of EU ceramic breeder development

• EU reference tritium breeding pebbles are produced using a melt-based process. • Process scale-up is underway to ensure the supply for ITER. • Biphasic materials show enhanced mechanical strength and stability. • Design relevant data were summarised in a material property handbook. • Crucial neutron irradiation experiments will complement the material qualification Biphasic ceramic pebbles, consisting of Li 4 SiO 4 and Li 2 TiO 3 , are being developed as the EU reference tritium breeding material for ITER and DEMO. A modified melt-based process for the pebble fabrication was established and optimised with regard to the process stability and yield. In parallel, the properties of pebbles and pebble beds were evaluated in dedicated experiments. These include investigations into e.g. the long-term thermal stability of pebbles, the thermomechanical properties of pebble beds, as well as the compatibility between the ceramics and the structural material. The available data was recently summarised and issued as the Material Property Handbook on Advanced Ceramic Breeders. During the EU DEMO conceptual phase, two main issues are proposed to be pursued in view of ITER. Irradiation campaigns are planned to ensure that the advanced ceramic breeder material is adequately tested and an extensive process scale-up with an increased production rate is foreseen to secure the supply for the ITER test blanket modules.

Effect of packing structure on anisotropic effective thermal conductivity of thin ceramic pebble bed

Helium cooled solid breeder blanket as an important blanket candidate of the Tokamak fusion reactor uses ceramic pebble bed for tritium breeding. Considering the poor effective thermal conductivity of the ceramic breeder pebble bed, thin structure of tritium breeder pebble bed is usually adopted in the blanket design. The container wall has a great influence on the thin pebble bed packing structure, especially for the assembly of mono-sized particles, and thin pebble bed will appear anisotropic effective thermal conductivity phenomenon. In this paper, thin ceramic pebble beds composed of 1 mm diameter Li4SiO4 particles are generated by the EDEM 2.7. The effective thermal conductivity of different thickness pebble beds in the three-dimensional directions are analyzed by three-dimensional thermal network method. It is observed that thin Li4SiO4 pebble bed showing anisotropic effective thermal conductivity under the practical design size. Normally, the effective thermal conductivity along the bed vertical direction is higher than the horizontal direction due to the gravity effect. As the thickness increases from 10 mm to 40 mm, the effective thermal conductivity of the pebble bed gradually increases.

Preparation of coated Li2TiO3 and Li4SiO4 pebbles by fluidized bed chemical vapor deposition for advanced tritium breeders

• Coated breeders were proposed to address safety hazard. • Protective coatings were successfully fabricated by FBCVD for the first time. • Deposition and interface reaction behaviors were revealed. Lithium-based ceramic tritium breeders would attack the blanket structural materials and produce brittle corrosion layers, causing potential safety hazard. Herein, coated breeders were proposed as advanced tritium breeders to address the issue. Carbon (C), titanium nitride (TiN) and titanium carbide (TiC) were investigated as inert protective shells because they show excellent stability and chemical compatibility with Li 2 TiO 3 and Li 4 SiO 4 . Moreover, for the first time, the C and TiN coated Li 2 TiO 3 and Li 4 SiO 4 pebbles were successfully fabricated by fluidized bed chemical vapor deposition (FBCVD) process, which would significantly decrease the brittle corrosion layers and increase the safety. The deposition behavior and interface reactions were systemically investigated.

MHD effects of partition plates on thermofluid performance of Indian variant LLCB TBM for ITER

Abstract Liquid metal Lead-Lithium (PbLi) flow channels in Indian LLCB (lead lithium cooled ceramic breeder) Test Blanket Module consist of a sequence of parallel channels which are fed from a common inlet header. Thin radial-toroidal vertical plates known as partition plates at the central plane perpendicular to the applied magnetic field are primarily envisaged to enhance the mechanical strength of high aspect ratio electrically coupled parallel poloidal channels. In that case the partition plates are acting as an internal Hartmann wall and they have significant effects on velocity profile, pressure drop and heat transfer characteristics at ITER relevant applied magnetic field (B0 = 4 T). Three dimensional numerical simulation is performed in a model of full scale LLCB variant to study the effects of these partition plates on various flow field variables. Axial velocity near the thin partition wall and side wall jet structures are found to get significantly. Magneto-hydraulic pressure drop and redistribution of flow rates due to the presence of partition plates are assessed by comparing the results with no partition wall. Thermofluid simulation is carried out to obtain the temperature distribution of various structures for a given heat load profile based on neutronic estimation. The effects of buoyancy force on thermofluid flow profile in channels with high temperature gradient are also presented.

An in-pile experimental loop for the irradiation of tritium breeding ceramics in China Mianyang research reactor (CMRR)

• An in-pile irradiation loop has been established in CMRR for in-situ tritium extraction study. • It allows loading or unloading the breeder pellets online by a special designed pneumatic system. • Several typical experiments have been performed to demonstrate its capability. A new in-pile experimental loop has been established in CMRR for ceramic tritium breeder irradiation. The most typical of this loop is the capability of online reloading of irradiated samples. The irradiation temperature in the pellets bed can be controlled within the range of 300～750 °C, with a spatial inhomogeneity of ±10 °C. Up to now, a series of irradiation experiments have been carried out on three different breeder materials, i.e. Li 4 SiO 4 (7.5 % 6 Li), Li 4 SiO 4 (93 % 6 Li) and Li 2 TiO 3 (7.5 % 6 Li) pellets, indicating the stability of the facility.

Mechanical assessment of first wall / blanket for DEMO

As a critical part of a breeding blanket of fusion reactor or DEMO, it is a common method to design and optimize the first wall separately before the mechanical evaluation of the fully integrated blanket. To understand the difference between the simplified first wall models and the fully integrated first wall/blanket model, the differences of the mechanical behavior between the simplified and integrated models are studied in this paper. The results provide a reference for the conceptual design and optimization of the first wall. Based on the HCCB (helium-cooled ceramic breeder) DEMO blanket model, the mechanical analysis of the integrated model was carried out, and the relatively accurate mechanical characteristics of the first wall for M-type and C-type damage were obtained. Considering the results of the integrated model as a basis, the mechanical analysis of the M-type and C-type for the 2 typical simplified model was carried out. For the separate first wall model, the maximum differences of M-type and C-type are -19% and -17%, respectively. For the most simplified model, the maximum differences of M-type are -17%, the differences of C-type are seriously deviated from the results of the integrated model. The results indicate that one quarter first wall model with reasonable constraints is closer to the results of the fully integrated FW/blanket model. The analysis results of one quarter first wall model can be used as a reference for the conceptual design of the first wall and can be utilized to optimize the first wall design based on the stress limits for prevention of the M-type and C-type damage modes.

Effects of bed dimension, friction coefficient and pebble size distribution on the packing structures of the pebble bed for solid tritium breeder blanket

• Effects of bed width on the packing structures of pebble bed were investigated by DEM simulation. • Effects of pebble properties (friction coefficient between pebbles, pebble size distribution and pebble density) on the packing fraction of pebble bed were simulated and analyzed. • Packing structures of both the tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) and the neutron multiplier (Be, Be 12 Ti) pebble beds were simulated and analyzed by DEM. In solid tritium breeder blanket, the packing structures of the tritium breeder pebble bed and the neutron multiplier pebble bed are very important to analyze the tritium breeder ratio and the heat and mass transfer process in the pebble bed. In this study, the numerous simulations were conducted to investigated the effects of bed dimension, friction coefficient between pebbles, pebble size distribution and pebble density on the packing structures of the ceramic tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) pebble beds and the neutron multiplier (Be, Be 12 Ti) pebble bed by discrete element method (DEM), respectively. The simulation results reveal that the pebble bed dimension, friction coefficient and pebble size distribution have a significant impact on the pebble bed packing fraction. As the width of the pebble bed container increases, the average packing fraction of the pebble bed gradually increases, meanwhile the influence of volume fraction of the wall effect gradually decreases. In addition, a lower friction coefficient between pebbles (smoother surface of the tritium breeder and the neutron multiplier pebbles) and wider pebble size distribution will result in a higher packing fraction of the pebble bed. However, the material density of the ceramic tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) pebbles and the neutron multiplier (Be, Be 12 Ti) pebble seems to have no obvious effect on the packing fraction of the wider pebble bed for the solid tritium breeder blanket.

Effect of packing structure on mechanical properties of the pebble beds and the probability of particle crushing

Abstract The tritium ceramic breeder pebble bed, used for tritium breeding and in nuclear fusion reactor, is a key constituent of helium cooled solid breeder blankets. The mechanical properties of the pebble beds, along with the integrity of the particles are very important for the extraction of tritium and safe operation of the solid blanket. In this paper, analysis is carried out on the effect of packing structure, with varying bed thicknesses and particle smoothness, on mechanical properties and the probability of particle crushing. Pebble beds are randomly packed by a single type of Li 4 SiO 4 particle, 1 mm in diameter. It is found that when the thickness of the Li 4 SiO 4 pebble bed (ranging from 15 mm to 25 mm) becomes thinner, or the coefficient of friction (ranging from 0.1 to 0.6) increases, the strain of the pebble bed and the probability of particle crushing increases under the same mechanical load. Pebble beds with different packing structures have the same probability distribution function of normalized contact forces. The results could serve to characterize the tritium pebble beds and to optimize the design of helium cooled solid breeder blankets.

Thermal dynamic analyses of the primary heat transfer system for the WCCB blanket of CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Testing Reactor (CFETR). The Primary Heat Transfer System (PHTS) of the WCCB blanket has been updated to fulfill the requirement of operating under 200 MW∼1.5 GW of fusion power based on the lasted CFETR design. The PHTS has two independent Cooling Systems (CSs). CS1 provides coolant for the First Wall (FW) and CS2 for Breeding Zones (BZs). In this paper, the PHTS design is optimized to obtain more uniform flow distribution among blanket sectors. Based on the optimized PHTS, plasma pulse operation and the Multifaceted Asymmetric Radiation From the Edge (MARFE) transient are simulated using RELAP5 to study the thermal dynamic behavior of the WCCB PHTS. The results show that CS1 and CS2 have different thermal dynamic behavior during plasma pulses due to different thermal inertia. Different schemes are considered for CS1 and CS2 to couple with the Power Conversion System (PCS). As for the MARFE, thermal hydraulic parameters of CS1 can return to steady state soon after the transient. The WCCB PHTS design is preliminarily evaluated as reliable according to thermal dynamic analyses.

Progress of water cooled ceramic breeder test blanket module system

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.

Effect of irradiation on the stability of the corrosion layer produced in EUROFER by contact with lithium ceramics

This paper is concerned with the identification of the corrosive phase formed on the surface of low-activation steel Eurofer by contact with a ceramic bleeder and the effect of irradiation on the stability of the phase.Tests were carried out at 550°C for 164 hours in a mixture of He/H2O (99.8/0.2 %vol) for simulating the purge gas. The results show an important interaction of the steel with the lithium ceramic breeder. The corrosion layer was characterized by scanning electron microscopy, x-ray diffraction, secondary ion mass spectroscopy and Mössbauer spectroscopy to verify the oxides formed during the corrosion test and the modification of the containing species after being subjected to different doses of ionizing radiation and / or deuterium absorption. The relationship of the different iron states and high energy traps for hydrogen isotopes (deuterium) that could be generated in a corrosion layer during reactor operation was also established.The results indicates that ionizing radiation up to the studied doses (200 MGy) does not produce additional defects to those generated by the transformation of hematite into non-stoichiometric wüstite, which is the fundamental compound of the corrosion layer by environmental deuterium.

Development of tritium dynamic transport analysis tool for tritium breeding blanket system using Modelica

Demonstration of the engineering feasibility of tritium breeding, including tritium generation, tritium extraction, tritium control and tritium safety is one of the main objectives of ITER Testing Blanket Module (TBM) Program. As one of TBM concepts, the tritium transport assessment of the China Helium Cooled Ceramic Breeder TBM and its ancillary systems (called test blanket system, TBS) is absolutely vital to understand and analyze the dynamic tritium behaviors in the system in detail according to the ITER operation scenario. Considering the requirements of the tritium transport analysis for the HCCB TBS, an integrated dynamic tritium transport analysis tool has been developed by the system simulation language “Modelica”. The multiple dynamic physics effects are considered in the tool, including the aspects impacting the tritium transport, simplified thermal-hydraulics considering only the linear loss and local loss in fluid, 1-d thermal conduction in solid, hydrogen isotope transport in fluid, 1-d tritium permeation in solid, tritium solution and recombination on solid surfaces, tritium desorption from ceramic breeder etc. A simplified model of the HCCB TBS is modeled to verify the functions and accuracy of this tool. The results, such as dynamic change of tritium concentration/partial pressure in fluids, dynamic change of tritium inventory in solids, tritium permeation fluxes etc., have been obtained, which are preliminary verified at various aspects by comparing with analytical and hand calculations.

Interface design of the blanket system for CFETR

The China Fusion Engineering Test Reactor (CFETR) aims to bridge the gaps between fusion experimental reactor ITER and demonstration fusion power reactor. Blanket, one of the key components of the CFETR, is of great importance to achieve the functions of tritium breeding, heat extraction and radiation shielding. As for the next device for fusion energy research in China, two blanket candidate options have been developing in parallel to decrease the risk of both physical and technical issues. Apart from the research and design work on the breeding blanket module itself, the interfaces definition and integration of the blanket system is vital, particularly due to the system interdependencies. The interface design of the blanket system is a challenge due to several factors, such as the thermal expansion, the electromagnetic loads, the feasibility of remote maintenance. In this paper, a preliminary interface definition and integration design of the CFETR blanket system has been presented, based on the design with R = 7.2 m and a = 2.2 m. The purpose is to complete the interface definition and integrated design of the blanket and other components or systems. The vacuum vessel has 16 sectors, and each sector has two inboard and three outboard blanket segments. The blanket segment will be installed and removed by remote maintenance tools through the upper ports. The work addresses the interfaces between the blanket and other system, such as diagnostic system and heating system. Preliminary assessment of the impact on tritium breeding ratio (TBR) for the candidate water cooled ceramic breeder blanket integration scheme has been presented highlighting the influence of the auxiliary interfaces system to TBR.