Blanket Materials Research Articles

Removal of bond sodium from Fermi-1 blanket assemblies using a melt-drain-evaporate process

A series of runs was performed to demonstrate the removal of bond sodium from full-length unirradiated Fermi-1 radial blanket elements and an entire radial blanket assembly using a melt-drain-evaporate process. The runs were performed in an inert atmosphere glovebox, where the blanket material was breached and loaded into a retort assembly. It was then subjected to elevated temperature and reduced pressure to melt, drain, and evaporate the bond sodium, while simultaneously collecting the sodium distillate in a crucible at the base of the assembly. After each run, the depleted uranium alloy slugs in every element slid out of its cladding, providing the first indication that the blanket material was effectually devoid of sodium metal. Entire columns of depleted uranium alloy slugs and associated cladding from select treated elements were separately subjected to quantitative analyses for residual sodium metal on the element surfaces, yielding removal efficiencies of ≥99.9998%. The remaining uranium alloy slugs and cladding segments from the series of runs were piece-wise immersed in ethanol to react with any residual sodium and provide a visual means of its presence or absence through the corresponding formation of hydrogen gas bubbles or lack thereof. Accordingly, the predominant absence of sodium reactivity on the balance of treated Fermi-1 blanket material was observed. The successful outcome of this demonstration provides a path forward for the treatment and disposal of 34 metric tons heavy metal (MTHM) of irradiated sodium-bonded blanket material from the decommissioned Fermi-1 reactor.

Preparation of Design Charts for Estimation of the Length of an Upstream Impervious Blanket in a Homogenous Earth Dam

Earth dams are constructed mainly from soil. A homogenous earth dam is composed of only one material. The seepage through such dams is quite high. Upstream impervious blanket is one of the methods used to control seepage through the dam foundations. Bennet's method is one of the commonly used methods to design an impervious upstream blanket. Design charts are developed relating the length of blanket, total reservoir head, total base width of the dam (excluding downstream drainage), the coefficient of permeability of the blanket material, blanket thickness, foundation thickness, and coefficient of permeability of the foundation soil, based on the equations governing the Bennet's method for a homogenous earth dam with a blanket of uniform thickness. The length of the upstream impervious blanket can be determined by using the developed charts. The length of the blanket is inversely proportional to the coefficient of permeability of the blanket material. The length of blanket is directly proportional to the total reservoir head, total base width of the dam (excluding downstream drainage), blanket thickness, foundation thickness, and coefficient of permeability of the foundation soil.

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.

A DFT + U study on diffusion and aggregation behavior of He atoms in Li2TiO3

Future fusion power plants are designed based on the deuterium-tritium fusion reactions. Yet, as an element with a half-life of 12.5 years, tritium cannot be obtained in large quantities from nature. To meet future demand, an artificial nuclear reaction is needed to use tritium breeder materials to produce tritium. At present, Li2TiO3 is the most promising tritium breeder blanket material in fusion reactor. The existing experimental results show that helium irradiation will change the crystal structure of Li2TiO3 and cause embrittlement failure. In order to study the microscopic helium embrittlement mechanism of Li2TiO3, the diffusion and aggregation behavior of helium atoms in Li2TiO3 bulk was calculated by first principles. The reasons for the formation of the most stable interstitial configuration of helium atoms and helium clusters in Li2TiO3 bulk are explained by lattice vibrational theory and electronic structure analysis. At the same time, the relatively new research results in this work are mainly as follows: In Li2TiO3 bulk, the three spatial inversion symmetries Li vacancy helium atoms tend to diffuse from all Li vacancies to Li2 vacancies. Meanwhile, the Li2 vacancy is the most likely site for helium clusters to nucleate. And helium atoms and the helium clusters tend to arrange and grow along the plane of the Li layer.

Tritium release behavior from neutron-irradiated FLiNaK mixed with Ti powder

• FLiNaK mixed with Ti powder was irradiated by mainly thermal neutrons. • Amounts of TF, T2 and T2O released from the FLiNaK by heating were investigated. • Chemical form ratio of released tritium was approximately TF:T2:T2O = 5:70:25. • Approximately 70% of tritium was released as T2O from the FLiNaK exposed to air. • Ti powder mixed to the FLiNaK contributed to the suppression of tritium release. FLiNaK is used as a simulant of FLiBe and FLiNaBe which are promising candidates of liquid blanket material in fusion reactors. In this study, to obtain useful knowledge for tritium release behavior from molten salts such as FLiBe and FLiNaBe, tritium release experiments were carried out for neutron-irradiated FLiNaK containing Ti powder. The solid-state samples were irradiated by mainly thermal neutrons at Kyoto University Research Reactor, and tritium release behavior as TF and T 2 by heating in Ar gas flow was observed and additionally total amount of released T 2 O was obtained. Most of the tritium generated inside the FLiNaK was released to the gas phase even when the FLiNaK was in a solid state at a temperature below its melting point. From the molten FLiNaK unintentionally exposed to air, approximately 70% of tritium was released as tritiated water vapor. The amount of TF released from the molten FLiNaK was considerably less than that of T 2 . This result is consistent with the tendency of tritium release from neutron-irradiated FLiNaBe in our previous study. The ratio of chemical form was approximately TF:T 2 :T 2 O = 5:70:25. It was experimentally confirmed that the Ti powder added to the salt contributes to the suppression of tritium release from the molten salt.

Tritium Breeding Ratio Evaluation of Solid Breeder Concepts for the FESS-FNSF

This paper presents a parametric study of the Fusion Energy System Studies-Fusion Nuclear Science Facility’s (FNSF’s) tritium breeding performance for several solid breeder concepts, neutron multiplying materials, and blanket materials, assuming volume fractions based on the most recent FNSF design as a realistically representative fusion facility. In this study, we initially surveyed the tritium breeding ratio (TBR) of several solid breeder concepts by employing a simplified but efficient one-dimensional (1-D) infinite cylinder reduced-order model (ROM). Parametric studies were performed with the ROMs for the full range of breeder-to-multiplier ratios to identify the optimum mixture compositions for each breeder type that would lead to a maximum TBR. These optimized breeder-multiplier combinations were then homogenized with FNSF blanket component materials to estimate their impacts on the TBR. Subsequently, as a validation step for the optimal designs, TBR calculations were performed using a more realistic modified 1-D ROM with inner and outer breeding regions, as well as with a fully detailed 22.5-deg three-dimensional (3-D) sector of the FNSF to assess the impact of geometry details on the TBR. The differences between the two 1-D models were negligible, while the ROMs were able to correctly predict trends and identify the maximum and minimum TBR cases, as well as show consistent biases relative to the results produced by the full 3-D, 22.5-deg sector for specific breeder/multiplier combinations. Solid breeder concepts such as , , and outperformed all others in this study in terms of TBR performance when combined with all the neutron multiplier materials selected. An underlying goal of this study was to develop and improve rapid and reliable ROMs to aid designers during parametric optimizations of highly complex and computationally expensive fusion models.

Analysis of continuous casting system for casting U-6%wt Zr slug

Uranium-rich U − Zr − Mo and U − Zr − Nb metal alloys are considered to be good alternative fuels for fast reactors. U-6 wt% Zr is one of the candidate blanket materials used in these fuel pins. In this study, U − Zr alloy samples were prepared using continuous casting method. Preparation of U-Zr alloy slugs through continuous casting process needs optimization of processing parameters such as, melt temperature, casting speed, cooling water flow rate in the cooling jacket. To arrive at the possible casting parameters and to reduce the number of experimental runs first computational analysis of the cast slug was carried out. Using these simulated results experimental runs were performed and the prepared slugs were further characterized by X-ray diffraction (XRD), scanning electron microscope (SEM) equipped with energy dispersive spectroscope (EDS).

Relevance of Tritium Breeder Irradiation Testing in a Fusion Prototypic Neutron Source

A fusion prototypic neutron source (FPNS) will play a fundamental role in testing and developing fusion materials that can withstand the harsh fusion environment. A thorough assessment is conducted to determine whether blanket materials would benefit from FPNS testing. Because of the cross sections for the various blanket materials, lithium-containing materials can generally be tested adequately using existing fission irradiation capabilities instead of waiting for or relying solely on future FPNS experiments. However, non-lithium blanket materials, such as beryllium multipliers and flow channel inserts, should be tested under a fusion neutron spectrum.

A benchmark study of large-angle neutron scattering cross section of tungsten using two shadow bars technique at 14 MeV

In fusion reactor design, neutron leaks intensively from blanket material through a gap. In this streaming phenomenon, backscattering cross section is known to be very crucial. In the present study, the author's team carried out a new experiment for benchmarking the large angle scattering cross section of tungsten using a DT neutron source of OKTAVIAN facility, Osaka University, Japan. Tungsten-containing material is under consideration as the radiation shield in a fusion reactor. The experimental geometry consists of a DT neutron source, two shadow bars, niobium foil, and a tungsten target. Four irradiations were performed at a neutron energy of 14 MeV using DT neutrons to extract only the contribution of large angle scattering cross section. By using two shadow bars, room return contribution was effectively suppressed. Consequently, only backscattering neutrons were measured by using a niobium foil. In the present benchmark study, obtained experimental data were compared with numerical calculations by MCNP6 using various nuclear data libraries, including JENDL-4.0, JENDL-5, JEFF-3.3, and ENDF/B-VIII.

Tritium Breeding Performance Analysis of HCLL Blanket Fusion Reactor Employing Vanadium Alloy (V-5Cr-5Ti) as First Wall Material

Neutronic analysis in the HCLL blanket module has been established, and the calculation was performed by the ITER team, including the first wall (FW). In this study, seven materials have been investigated for FW material by considering characteristics such as high neutron fluence capability, low degradation, under irradiation, and high compatibility for blanket material. A three-dimensional configuration simulated in MCNP5 program codes was performed to investigate the neutronic performance and radiation damage effect. Employing seven candidates are vanadium carbide (VC), titanium carbide (TiC), vanadium alloy (V-5Cr-5Ti), graphite (C), tungsten alloy (W-CuCrZr), ceramic alloy (SiC), and HT-9 to study optimization of FW materials configurated in the HCLL blanket module. This novelty study concludes that vanadium alloy (V-5Cr-5Ti) is becoming a promising material candidate. This alloy has the highest number of neutronic performing for 1.27 TBR and 1.26 in multiplication energy factor in all investigations. Meanwhile, the amount of atomic displacement, hydrogen, and helium production are around 22.31 appm, 765.55 appm, and 281.57 appm, respectively. Even though vanadium alloy has a reasonably high radiation damage effect, it is still tolerable compared to several thresholds of DPA. So, it is considered excellent material for FW. Nevertheless, this alloy can replace after 13.45 years for radiation damage.

Solubility of Bi in Li–Pb Eutectic Alloy Between 508 and 623 K

Liquid lithium–lead (Li–Pb) eutectic alloy is one of the candidate liquid tritium breeding materials of fusion blanket and it contains bismuth (Bi) as an initial impurity or transmutation product from lead. The saturated solubility of Bi, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm), in Li–Pb was measured in the temperature range from 508 to 623 K. The measured results were plotted on the graph with <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm) versus 1/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T$ </tex-math></inline-formula> . From these plots, a positive temperature dependence, where the solubility decreases as the temperature is lowered, is confirmed. The equation of the solubility of Bi in Li–Pb was obtained, too. In addition, the dissolution enthalpy of Bi in Li–Pb, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\Delta H_{\mathrm {sol}}$ </tex-math></inline-formula> , was quantified from this equation. The precipitation morphology of Bi in Li–Pb was also investigated. X-ray diffraction was used to analyze the precipitates collected from the surface of the liquid phase. As a result, the peak pattern of Li3Bi was found from the precipitate. Moreover, the amount of polonium produced when the cold trap is introduced to a Li–Pb loop of a fusion reactor was predicted from the solubility obtained in this study.

Tritium Transport Modeling at the Pore Scale in Ceramic Breeder Materials Using TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders (CCBs) are dense materials that can maintain a high tritium breeding ratio while promoting tritium release because of highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport [i.e., the Tritium Migration Analysis Program (TMAP8)], which is itself built on top of the finite-element multiphysics framework multiphysics object-oriented simulation environment (MOOSE). This study demonstrates some of TMAP8’s capabilities and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

Passive cooling in Aerogel-Based insulation systems for liquid hydrogen upper stage launch vehicle tanks

Spray-On Foam Insulation (SOFI) is typically used to protect the windward-facing side of flight tanks for space launch vehicle liquid hydrogen (LH2) stages. SOFI is an excellent insulator in ambient pressure environments, however, its performance pales in comparison to reflective-type systems such as Multi-Layer Insulation in the thermal radiation dominated vacuum environment of space. If a windward insulation system employing radiation shields could be devised to replace or supplement SOFI, the on-orbit heat load could be drastically reduced, and the residual propellants could then be used to facilitate secondary missions. Such insulation systems, utilizing aerogel blanket insulation, have been explored by the Cryogenics Test Laboratory at NASA Kennedy Space Center. Owing to its nano-porous structure, aerogel is an excellent adsorbent as well as an insulator at cryogenic temperatures and will readily uptake condensable background gasses such as air when protecting a surface near LH2 temperatures. When an adsorbed blanket is rapidly exposed to a vacuum, such as during the ascent of a rocket, it will release the background gas, producing a passive cooling effect that could potentially reduce or eliminate the heat load on the propellant tank for a time. This work discusses the setup and results of a test program carried out to determine the effectiveness and impact of this cooling effect as it pertains to LH2 upper stage tanks. A vertical-cylindrical liquid nitrogen (LN2) calorimeter test instrument acted as the upper stage analog. The cold-mass was wrapped with two layers of 10 mm thick aerogel blanket material, with and without aluminum heat shields depending on the test, and interspersed with numerous thermocouple temperature sensors. Gaseous argon was used as a stand-in for air as the condensable background gas. The test procedure mimicked vehicle cryo-loading/stabilization, and ascent phases, with the ascent phase accomplished using parallel vacuum pumps. Results show that the desorption cooling effect was enough to not only eliminate the heat load on the LN2 cold-mass entirely, but further refrigerated the liquid below its boiling point.

Influence of sintering atmosphere on phase, microstructure and mechanical properties of Li4Si0.7Ti0.3O4 tritium breeding ceramics

Li4Si1–xTixO4 ceramic solid solution was considered as a promising tritium breeding material for fusion reactor blanket. In this work, the effects of the sintering atmosphere on the phase composition, microstructure, and mechanical properties of Li4Si0.7Ti0.3O4 ceramic pebbles were investigated for the first time, aiming to explore the optimal sintering atmosphere for this solid solution. The results show that compared with air and argon atmosphere, the negative pressure of vacuum atmosphere was more favorable for Ti to enter the Li4SiO4 crystal structure as a substitute for Si atom to form a solid solution. More importantly, sintering in argon-vacuum atmosphere can ensure the formation of solid solution while fully exerting the liquid phase sintering effect. The crushing load of the Li4Si0.7Ti0.3O4 ceramic pebbles sintered at 800 °C in argon-vacuum atmosphere reaches 30.3 ± 3.2 N. In general, the appropriate sintering atmosphere was of great significance for obtaining Li4Si0.7Ti0.3O4 solid solution to meet the performance requirements of tritium breeding blanket.

Study of the Effect of Two Phases in Li4SiO4-Li2SiO3 Ceramics on the Strength and Thermophysical Parameters.

The paper studies the effect of Li2SiO3/Li4SiO4 phase formation in lithium-containing ceramics on the strength and thermophysical characteristics of lithium-containing ceramics, which have great prospects for use as blanket materials for tritium propagation. During the phase composition analysis of the studied ceramics using the X-ray diffraction method, it was found that an increase in the lithium component during synthesis leads to the formation of an additional orthorhombic Li2SiO3 phase, and the main phase in ceramics is the monoclinic Li4SiO4 phase. An analysis of the morphological features of the synthesized ceramics showed that an increase in the Li2SiO3 impurity phase leads to ceramic densification and the formation of impurity grains near grain boundaries and joints. During determination of the strength characteristics of the studied ceramics, a positive effect of an increase in the Li2SiO3 impurity phase and dimensional factors on the strengthening and increase in the resistance to external influences during compression of ceramics was established. During tests for resistance to long-term thermal heating, it was found that for two-phase ceramics, the decrease in strength and thermophysical characteristics after 500 h of annealing was less than 5%, which indicates a high resistance and stability of these ceramics in comparison with single-phase orthosilicate ceramics.

Study of Structural and Strength Changes in Lithium-Containing Ceramics-Potential Blanket Materials for Nuclear Power, Subjected to High-Dose Proton Irradiation.

The paper considers the hydrogenation processes in Li2TiO3 ceramics under irradiation with protons with an energy of 500 keV and fluences of 1 × 1010–5 × 1017 ion/cm2. The choice of the type of irradiation, as well as the irradiation fluences, is based on the possibilities of modeling hydrogenation processes and studying the kinetics of structural changes caused by the accumulation of radiation damage. The choice of Li2TiO3 ceramics as objects of research is due to their prospects for using as blanket materials of thermonuclear reactors for the tritium production and accumulation. It was found that the formation of point defects and their subsequent evolution associated with the formation of complex compounds and the filling of pores, followed by the formation of gas-filled bubbles, the presence of which leads to a decrease in crack resistance and resistance to destruction of the near-surface layer. Based on the data on structural changes and evolution of the crystal lattice parameters, its swelling, a description of the destruction processes associated with hydrogenation in Li2TiO3 ceramics was proposed. Also, during the studies, it was found that at irradiation fluences above 1 × 1017 ion/cm2, the appearance of impurity inclusions characteristic of the TiO2 phase was observed, the presence of which indicates the crystal lattice destruction processes because of accumulation of radiation damage and deformations caused by them. Critical doses are established at which there is a sharp deterioration in strength and crack resistance, reflecting the resistance of ceramics to mechanical external influences.

Development of reduced activation ferritic/martensitic steels in China

Reduced-activation ferritic-martensitic (RAFM) steel is considered as the primary candidate structural material for blankets of ITER, CFETR and DEMO. Several RAFMs have been developed in China (CN-RAFMs) such as CLAM, CLF-1, modified RAFMs and oxide dispersion strengthened RAFMs (ODS-RAFMs). The mechanical properties, irradiation behaviors, additive manufacturing and joining technologies of structural materials have been comprehensively studied. Qualifications of CN-RAFMs are on-going for ITER-TBM fabrication, with emphasis on material property databases and standardization. A new neutron irradiation program is being carried out for a series of fusion materials e.g. RAFMs and ODS-RAFMs with doses up to 40 dpa. Joining technologies (e.g. electron beam welding, tungsten inert gas welding, laser beam welding and diffusion welding, etc.) and advanced additive manufacturing are performed to overcome the big challenges of the blanket fabrication before ITER, CFETR and DEMO. The databases and standards for CLAM and CLF-1 were established respectively, and the specification and qualification of them are underway for ITER-TBM.In this paper, the latest development and strategy on fusion energy in China are briefly introduced, and the progresses of R&D activities on CN-RAFMs aiming for the engineering applications are reviewed.

Synergies between H, He and radiation damage in dual and triple ion irradiation of candidate fusion blanket materials

Three ferritic/martensitic alloys were studied to understand the synergistic effect between single ion beam (Fe2+), dual ion beam (Fe2++He2+ and Fe2++H+), and triple ion beam (Fe2++He2++H+) irradiations on cavity evolution. A commercial alloy, F82H, a castable nanostructured alloy, CNA3, and a high purity model alloy, Fe8Cr2W, were irradiated at 400°C to 600°C to a damage level of 50 dpa at a damage rate of 1 × 10−3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field transmission electron microscopy and high-angle annular dark-field scanning transmission electron microscopy was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium resulted in an increased cavity number density and maximum cavity size, producing an increase in swelling over that from helium injection alone. Swelling in F82H appears to peak between 450°C and 500°C. At 600°C, swelling was minimal and cavities of high density and small size were confined to grain boundaries and dislocations while at 400°C, swelling is also low with a nearly homogeneous, high density, distribution of very small cavities throughout. Swelling was least in the commercial alloy F82H due to the high sink strength. The CNA3 alloy underwent dissolution of precipitates that lowered the sink strength and resulted in higher swelling than F82H, but less than the model alloy. Electron energy loss spectroscopy (EELS) elemental mapping revealed hydrogen forming a halo-like structure about the periphery of the cavities and helium residing within the cavities themselves. This observation suggests that hydrogen reduces the surface energy of helium-filled cavities which results in both increased cavity number density and cavity size in triple beam irradiation over dual beam irradiation.

A Model to Simulate Gas Dissolution into/through Metals and Its Application to Deuterium in a 316L Steel Chamber with Pb-Li in a Quasi-2D Geometry

Liquid lead-lithium in eutectic proportions (Pb-Li) is a candidate material for Breeding Blankets (BBs) in future Fusion Power Plants (FPP). BB design depends on the diffusivity and Sieverts’ constant (solubility) of tritium in this alloy, but literature reports a large scattering of measurements for these values. A model was developed to address one possible source of this scattering in static experiments, i.e., non-negligible loss of hydrogen gas through steel walls of containers. This model simulates the dissolution of gases into, and their diffusion through, metallic barriers for diffusivity and Sieverts’ constant as inputs. When implemented, it can be used to compute the pressure decrease in a metallic chamber, and comparison of simulated curves with experimental ones allows for estimates of the diffusivity and Sieverts’ constant. This approach was used to estimate these coefficients for deuterium in stainless steel, using experiments performed with a 316L steel chamber from an existing facility (the Vacuum Sieve Tray setup) and simulations in a quasi-2D representation of this chamber. This validated the model, which was then used to simulate the chamber containing Pb-Li, as a means of planning for future experiments.

Li4SiO4 breeder pebbles fabrication by a sol-gel supported drip casting method

Lithium orthosilicate (Li4SiO4) is widely considered a potential candidate as tritium breeding material for blankets in thermonuclear fusion reactors. In this study, a low-temperature methodology based on the drip casting forming technique was proposed for Li4SiO4 pebbles fabrication. The hydrolytic sol-gel method (starting from tetraethyl orthosilicate and lithium hydroxide) was applied to synthesize Li4SiO4 precursor powders, which were used to prepare an aqueous suspension for dripping in calcium chloride/lithium acetate solution. Spherical Li4SiO4 pebbles were obtained after sintering (900-950°C) and exhibited regular roundness, high phase purity, and no macroscopic defects. A high density (> 70 % of the theoretical Li4SiO4 density) and a good crush load (35 N) were achieved when cellulose fibers were added to the precursor suspension and the sintering step was conducted up to 950°C. Results indicate that the produced Li4SiO4 pebbles can be suitable for the use in breeding blankets in future fusion reactors.