Solid Breeder Blanket Research Articles

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.

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.

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.

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.

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.

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.

Deuterium permeation behavior of zirconium oxide coating after exposure to solid breeder pebbles

Tritium permeation barrier has been developed using ceramic coatings to mitigate tritium leakage through structural materials in most of the fusion reactor blanket concepts. In a solid breeder blanket concept, corrosion of the coating is a serious issue in addition to tritium permeation; however, the effects of corrosion on tritium permeation have been rarely researched. Here, we report deuterium permeation behaviors of zirconium oxide coatings prepared by metal organic decomposition after exposure to two kinds of solid breeder pebbles with different compositions, followed by deuterium permeation measurements. The deuterium permeation flux varied with the pebble composition: the coated sample after exposure to the pebbles of 30 mol% lithium metatitanate and 70 mol% lithium orthosilicate was smaller than that after exposure to the pebbles of 20 mol% lithium metatitanate and 80 mol% lithium orthosilicate. Because the thickness of the corrosion layer formed by the pebbles with 30 mol% lithium metatitanate and 70 mol% lithium orthosilicate was larger, the composition of the pebbles affected the thickness of the corrosion layer and the permeation behaviors rather than coating integrity.

Thermal desorption of tritium and helium from lithium ceramics Li2TiO3+5mol% TiO2 after neutron irradiation

Lithium-based ceramics exhibit good thermophysical properties, have low activation and chemical activity, and also release tritium well. These characteristics make lithium ceramics the best candidates for use as a functional material for a solid breeder blanket. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials can only be obtained from the results of reactor experiments. On the other hand, many important parameters can only be estimated in post-irradiation experiments (PIE). This paper describes studies of tritium and helium release in post-irradiation experiments (PIE) on the thermal desorption from ceramic pebbles Li2TiO3 + 5mol% TiO2 (with 96% enrichment on lithium-6 isotope), irradiated in the WWR-K reactor for 223 days. In PIE a peak of helium release was recorded for each sample in the temperature range of 1300–1500 K. An assumption was made in the paper to explain the nature of this helium release peaks from lithium ceramic pebbles.

One‐step fabrication of Li2TiO3 ceramic pebbles using pulsed YAG laser

AbstractA large amount of Li‐containing ceramic breeder pebbles is packed in the solid breeding blanket of a nuclear fusion reactor. Several pebble fabrication technologies have been proposed in previous studies, including wet process, emulsion method, extrusion spheronization, additive manufacturing, and melt process. However, a simple, energy‐effective, and scalable fabrication technology remains to be developed for the automated mass production and reprocessing of used radioactive pebbles post‐operation. Selective laser melting potentially enables the quick and automated fabrication of breeder pebbles. Herein, we employ a high‐power density pulse laser to produce ceramic breeder pebbles. A pulsed YAG laser was irradiated over a lithium metatitanate (Li2TiO3) powder bed in air, and the corresponding temperature was monitored using fiber‐type infrared pyrometers. Spherical Li2TiO3 pebbles were successfully fabricated in a single step with an average diameter of 0.78 ± 0.13 μm and the sintering density of 87.4% ± 5.6% (input power: 7.9 J/pulse). The irradiated Li2TiO3 powder melted and turned spherical under surface tension and rapidly solidified, resulting in uniaxial fine grains and a decrease in the degree of long‐range cation ordering.

Pore-scale simulation on flow and heat transfer characteristics in packed beds with internal heat sources at low Reynolds numbers

Packed beds with internal heat sources are widely used in various engineering applications. But the effects of the internal heat source on the flow and heat transfer in the packed beds received less attention. In this study, the flow and heat transfer characteristics in packed beds with internal heat sources were studied at low Reynolds numbers to assist the design of the solid breeding blanket in the fusion reactors. Through the pore-scale simulations, the local behaviors (including flow velocity, temperature and fluid properties) and macroscopic flow and heat transfer characteristics were analyzed. It was found that with the internal heat sources, the fluid properties vary significantly form the packing inlet to outlet (e.g. the density decreases by 34% and the viscosity increases by 31% in the most). Due to the variations of the fluid density and viscosity, the conventional Ergun and Blake-Kozeny-Macdonald equation do not predict the pressure drop in the packed bed well. Then a modified model by introducing a modified coefficient considering the variable properties is proposed to predict the pressure drop of the packed bed with internal heat sources. The average deviation between the modified pressure drop model and the pore-scale simulation results is only 1.5%. Then a modified heat transfer model was proposed to predict the heat transfer characteristics of the packed bed with internal heat sources at low Reynolds number flow (Re <1.8). The deviations of the predictions by the modified model are within ±30% from the simulation values.

Investigation and Analysis on Tritium Accumulation Phenomenon Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a major scientific project independently designed and developed by China, which will demonstrate the engineering feasibility of continuous large-scale fusion energy for safe and stable power generation. One of the most critical issues for fusion reactors is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.0, as it relates to the ability of the fusion reactor to fulfill the fuel cycle and continue stable operation. Moreover, the release and inventory of tritium is also a significant factor affecting the operation of CFETR. The temperature distribution inside the breeder blanket is precisely related to the tritium release. Therefore, it is particularly important for CFETR to optimize the temperature field inside the breeding blanket. It is significant that to ensure that the temperature of the mixed pebble bed zone reaches the limit of 900 °C as much as possible, which could maximize the tritium release efficiency. Simultaneously, it must make sure that other structural materials meet the allowable value of temperature. Tritium accumulation close to the cooling pipe is investigation, which will be adverse to tritium release. Hence, in order to solve this difficulty, the Be12 Ti tube was sleeved on the cooling pipe to increase the temperature of the mixed pebble bed zone near the cooling pipe. In this work, a water-cooled solid breeder blanket was proposed and the nuclear heat was calculated based on the 1.5 GW fusion power. The nuclear heat was used as the preprocessing data for thermal calculation carried out by the Monte Carlo N-Particle Transport Code (MCNP). With the given conditions, the thermal hydraulic was simulated with the commercial CFD code, the ANSYS CFX. In this way, a breeding blanket structure with a uniform temperature field of mixed bed zone close to 900 °C can be obtained. In addition, the selection of Be12 Ti tubes with the thickness of 0.525 cm can ensure a high TBR and increase the temperature around the cooling tubes to improve the tritium release efficiency. This work provides an important reference for the subsequent engineering design of CFETR breeding blanket.

Reaction kinetics of lithium extraction and a recycling process for lithium meta-titanate pebbles

ABSTRACT Fusion reactors require large amounts of lithium meta-titanate (Li2TiO3) pebbles as tritium breeding materials for Japanese water-cooled solid breeder blanket. The development of a recycling process for tritium-breeding materials is essential for the early realization of fusion reactors. This study was proposed to understand the kinetics of Li extraction from Li2TiO3 by inducing chemical reactions and applying microwave heating and establish a highly efficient recycling process for spent Li2TiO3 pebbles. The results of Li extraction experiments performed under various conditions clarified that this extraction reaction could be applied via a surface chemical reaction with kct = 1 − (1 − f)1/3. The apparent activation energy associated with this extraction reaction was evaluated to be 52.78 kJ/mol, which is in agreement with that of the chemical reaction-controlled mechanism. Additionally, the reaction order for the dissolution of Li2TiO3 in the HCl solution was first-order. The highly efficient recycling process of Li2TiO3 pebbles was suggested.

A new direct external recycling concept for the tritium fuel cycle of CFETR

Tritium self-sustaining operation, which is achieved by a successful closed tritium fuel cycle, is one of the key scientific objectives of China Fusion Engineering Test Reactor (CFETR). The initial tritium start-up inventory and the required TBR are two key parameters which directly affect the possibility of achieving tritium self-sufficiency. In this paper, a new purge gas scheme of He + 0.n% vol. D2 is proposed for tritium extraction of solid breeder blanket, which can avoid the existence of protium in the mixed gas. In this way, the isotope separation system (ISS) in outer fuel cycle (OFC) can be omitted and a new concept of direct external recycling (DER) has been first put forward for OFC, which aims to skip ISS-O and reduce the processing time of OFC; and the feasibility of DER is preliminarily analyzed from four aspects. On this basis, the influences of DER on both the initial start-up inventory and the required TBR have been preliminarily analyzed for different fueling efficiency and tritium burning fraction (TBF) products, and the calculation results are compared with those corresponding to the same direct internal recycling (DIR) fractions. The results indicate that the application of DER can reduce the demand for both the initial start-up inventory and the required TBR, which can improve the possibility of achieving tritium self-sufficiency appropriately. In particular, the amount of “start-up inventory saved by DER” is still attractive under the condition of high fueling efficiency and TBF, and it is in direct proportion to the fusion power. Both of these make DER technology even more attractive for DEMO reactor and future power plant (FPP). These studies can provide some valuable suggestions for the design and optimization of the tritium fuel cycle system of CFETR and FPP.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

Simulation and experimental analysis of purge gas flow characteristic for pebble bed

In a solid breeder blanket as a functional material, the lithium ceramics are kept in the form of a pebble bed. Helium is used as a purge gas, which flows through the pebble bed. The flow characteristics are essential in designing and running the breeding blanket efficiently, depending on the pebble bed arrangement. In the present study, a computational model of unitary pebble bed was engineered using DDPM-DEM (Dense Dispersed Phase Model-Discrete Element Method) to study the purge gas flow characteristics of the pebble bed. A similar packed bed was engineered using a porous media approach. The parameters which affect the flow characteristics are porosity, pressure distribution, pressure drop, and wall effect. Nitrogen and helium were used as purge gas to study the flow characteristics of a pebble bed. The simulations with the laminar and k-ɛ model were performed at different inlet velocities, and flow characteristics of a packed bed were observed. The comparison shows that the pressure drop simulation results agree with the Ergun equation's predicted pressure drop. Also, a previously developed experimental facility has been used to validate the simulation results. Artificial Neural Network (ANN) is used to predict the pressure drop as a complementary approach. DDPM-DEM analysis helps to predict the pressure drop which enables us to finalize the design parameters of canister, pebble size, finally the size of the pebble bed, packing fraction etc. Simulation results were also compared with the experimental results and ANN model, which were in good agreement.

Neutronics optimization study on the first wall design of CFETR for TBR enhancement

Tritium breeding capability is one of the most important objectives for fusion reactor design, and the tritium self-sufficiency has been set as a crucial target for the China Fusion Engineering Test Reactor (CFETR). The first wall (FW) is the first component of the blanket facing plasma, which bears high heat flux and high neutron current. Due to the neutrons entering into the blanket pass through the FW and then absorbed by the breeder materials, the FW design directly influences the blanket neutronics and tritium breeding performance. In this study, the helium-cooled solid breeder (HCSB) blanket of CFETR is selected as the reference design, and the impacts of the FW design parameters on tritium breeding performance have been studied comprehensively. A new conceptual mixed armor design and FW neutronics optimization guidelines have been proposed to enhance the tritium breeding ratio (TBR). Based on the mixed armor design and the optimization guidelines, the preliminarily optimized helium-cooled FW design shows about 9% TBR enhancement.

A Water cooled Lead Ceramic Breeder blanket for European DEMO

The development of a demonstration fusion power plant (DEMO) is widely considered a crucial step beyond ITER leading toward fusion electricity. In a viable DEMO plant two major functions have to be achieved, namely to breed tritium for fuel self-sufficiency and to extract high-grade heat to allow an efficient electricity production. The breeding blanket is the key component to ensure these two functions. The solid breeder blanket is considered in the design of several ITER Test blankets and DEMO breeding blankets. In Europe, the helium cooled solid breeder blanket, so-called Helium Cooled Pebble Bed (HCPB) blanket, is one of the two driver blanket candidates selected for the European DEMO. In spite of the considerable progress and achievements, several potential issues inherent to the choice of the neutron multiplier and coolant remain with this concept. To overcome these challenges, research activities on an innovative concept (a Water cooled Lead Ceramic Breeder - WLCB), in which Lead/Lead-alloy is used as neutron multiplier, ceramic pebbles as tritium breeder, pressurized sub-cooled water as coolant, are on-going. This concept avoids the potential safety issue (exothermic hydrogen-producing Be-water reaction or PbLi-water reaction in case of water leakage) when Beryllium/Beryllide is used as neutron multiplier or lithium lead is used tritium breeder in a water cooled blanket concept. The neutronics, thermal hydraulics and mechanical performance figures of the above-mentioned innovative blanket concept are presented. This work concludes with possible future work and outlook towards the development of innovative alternative blanket concepts for the European demonstration fusion power plants.

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.