Liquid LiPb Research Articles

Chemical and structural durability of α-Al2O3 and γ-LiAlO2 layers formed on ODS FeCrAl alloys in liquid lithium lead stirred flow

The protection performance of an α-Al2O3 layer formed on oxide dispersion-strengthened (ODS) FeCrAl alloys in liquid LiPb flow was investigated by means of the corrosion tests under stirred-flow condition at 873 K for 1000 h. The result of STEM/EELS analysis indicated the Li adsorption and diffusion into α-Al2O3 layer. The ODS FeCrAl alloys in-situ formed a durable γ-LiAlO2 layer on their bare surface in liquid LiPb. The results of micro scratch tests on the protective layers indicated that they revealed strong resistance against the exfoliation in a shearing direction after the immersion to liquid LiPb.

Atom Probe Tomography investigation of lithium uptake from liquid Pb-17Li by nanocrystalline/amorphous Al2O3 coatings

Atom probe tomography has been applied to study lithium-induced corrosion of Pulsed-Laser Deposition grown Al2O3, which is a candidate coating material for breeder blankets for nuclear fusion reactors. Coated Eurofer97 specimens were immersed in static Pb-17Li for 7000 h with samples extracted from the regions near the ceramic-substrate interface to determine the extent of Li penetration. Li-enriched regions, which are non-stoichiometric with respect to any known equilibrium phases predicted by the Li-Al-O phase diagram at 823 K, are reported for the first time in Al2O3. The nature and composition of the Li-enriched features were characterized with nanometre resolution. It was found that the Li content inside these features may increase to a maximum of 35 at%, with the adjacent Al2O3 matrix containing approximately 5–10 at% Li. Based on these results, the corrosion mechanism of amorphous Al2O3 is proposed to be Li substitution of Al atoms.

Stochastic computer experiments of the thermodynamic irreversibility of bulk nanobubbles in supersaturated and weak gas-liquid solutions.

Spontaneous gas-bubble nucleation in weak gas-liquid solutions has been a challenging topic in theory, experimentation, and computer simulations. In analogy with recent advances in crystallization and droplet formation studies, the diffusive-shielding stabilization and thermodynamic irreversibility of bulk nanobubble (bNB) mechanisms are revisited and deployed to characterize nucleation processes in a stochastic framework of computer experiments using the large-scale atomic/molecular massively parallel simulator code. Theoretical bases, assumptions, and limitations underlying the irreversibility hypothesis of bNBs, and their computational counterparts, are extensively described and illustrated. In essence, it is established that the irreversibility hypothesis can be numerically investigated by converging the system volume (due to the finiteness of interatomic forces) and the initial dissolved-gas concentration in the solution (due to the single-bNB limitation). Helium nucleation in liquid Pb17Li alloy is selected as a representative case study, where it exhibits typical characteristics of noble-gas/liquid-metal systems. The proposed framework lays down the bases on which the stability of gas-bNBs in weak and supersaturated gas-liquid solutions can be inferred and explained from a novel perspective. In essence, it stochastically marches toward a unique irreversible state along out-of-equilibrium nucleation/growth trajectories. Moreover, it does not attempt to characterize the interface or any interface-related properties, neither theoretically nor computationally. It was concluded that bNBs of a few tens of He-atoms are irreversible when dissolved-He concentrations in the weak gas-liquid solution are at least ∼50 and ∼105 molm-3 at 600 and 1000K (and ∼80 MPa), respectively, whereas classical molecular dynamics -estimated solubilities are at least two orders of magnitude smaller.

Continuous deuterium extraction from falling lithium-lead droplets in a vacuum

This study experimentally verifies continuous deuterium extraction efficiency from falling liquid lithium-lead (LiPb) droplets in a vacuum, as an alternative to tritium The obtained efficiencies, from five runs, were between 0.59 and 0.69. The corresponding dispersion coefficient of deuterium in LiPb droplets is (2.6 + 0.8 – 0.6) × 10–7 m2s–1. These results closely resemble those obtained in a proof-of-principle (POP) experiment conducted at Kyoto University in 2013. Consequently, the deuterium release from falling LiPb droplets in a vacuum can be considered a stable function. The experimental campaign was carried out using a dedicated setup constructed on the liquid metal test loop (Oroshhi-2) at the National Institute for Fusion Science. Tritium breeding in a liquid LiPb blanket is substituted by the deuterium dissolution in circulating LiPb. The extraction efficiency was determined by comparing the time-series deuterium flux, released from falling droplets, with simulation. An alternative method involving a before and after concentration comparison did not yield consistent results. This inconsistency can be attributed to issues with the concentration sensor, which experienced difficulties due to surface degradation. The experimental parameters, the LiPb flow rate, nozzle diameter, falling height, temperature, and duration period, were set to 0.1 to 0.3 l-min–1, 1.0 mm, 500 mm, 350 °C, and 9 h, respectively.

Fabrication and characterization of zirconium oxide coating on tubular steel by metal organic decomposition

In fusion reactor blankets, multifunctional ceramic coatings fabricated on structural materials have been investigated for tritium permeation reduction, electrical insulation, and corrosion protection. However, it has been challenging to manufacture a homogeneous coating on the complicated components in the blanket breeder, such as ducts. In terms of sustainable development, it is crucial to establish a ceramic coating technique on the inner walls of fuel pipes in liquid blanket systems. In this study, we applied the metal organic decomposition method to fabricate zirconium oxide coatings on 316L austenitic stainless steel tubes and characterized their properties as liquid tritium breeding pipes. Homogeneous zirconium oxide coatings with a thickness of 395 ± 12 nm have formed on the inner walls of SS316L tubes. The compatibility of the coated tubes with liquid lithium-lead was examined under static and flowing conditions at 550–600 °C for prolonged exposure of up to 2000 h. Under static conditions, ZrO2 coatings demonstrated exceptional compatibility with the liquid Li-Pb as they protected structural material from corrosion. At a flow rate of 0.16 m/s, the coatings were partly delaminated due to different thermal expansion coefficients of the coating and substrate and acceleration under a dynamic system. The coating thickness increased up to 650 ± 26 nm after exposure for 2000 h, which indicated ZrO2 coatings reacted with Li-Pb to form corrosion products. Besides, ZrO2 coatings achieved the insulating requirements of magnetohydrodynamic insulators as their electrical resistivities were on the order of 104–106 Ω m at the temperature range of 250–550 °C.

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.

Evaluation of the strength of FeCrAl alloy/surface Oxide film interface by micro double-notch shear test

The shear strength of the film formed by pre-oxidation treatment of FeCrAl(-ODS) alloys, likely to be used as pipework material for liquid LiPb breeder blankets, was evaluated by micro double-notch shear testing. Fracture generally occurred at the interface between the oxide film and the alloy substrate, but the pop- in the stress-strain curves was difficult to interpret, so all the test results in this study were compared to the full rupture strength. The rupture strength appears to be a function of the nature of the interface between the film and the alloy rather than the thickness of the film; for APMT and FeCrAl-ODS, the rupture strength was greater for FeCrAl-ODS.

Deuterium permeation and lithium‑lead corrosion behaviors of ceramic‑iron joint coating

In fusion reactor blanket design, the main challenges are to control the permeation of tritium and prevent the corrosion of the structural materials. This study focuses on investigating the behavior of ceramic‑iron joint coatings in terms of deuterium permeation and Li-Pb corrosion aiming to develop effective barriers against permeation while providing corrosion protection. The research involved fabricating ZrO2 and Fe2O3 ceramic coatings on F82H steel substrates using metal-organic decomposition followed by joining the coated samples with Fe foils using a hot press machine. Deuterium permeation measurements were conducted to assess the effectiveness of the coating, and liquid Li-Pb exposure tests in a flow environment were performed to evaluate its resistance against corrosion. The results showed that the coating reduced deuterium permeation flux by a factor of 2000 at 400 °C and 7000 at 550 °C compared to uncoated samples suggesting an improved crystallinity or grain growth without degradation due to the joining process. Furthermore, the corrosion layers formed on the surface of the Fe layer after the flowing Li-Pb exposure test. Although this layer gradually decreased in thickness over time while not significantly affecting the underlying Fe layer's thickness. This observation indicates that the corrosion layer performed mitigating the corrosion rate of the Fe layer. This paper aims to demonstrate how ceramic‑iron joint coatings can be compatible with the advanced fusion reactors.

Axial Concentration of Dissolved Deuterium by Dispersion in a LiPb Experimental Loop

The axial concentration of deuterium by dispersion in a circulating liquid lithium-lead (LiPb) loop was analyzed and experimentally verified. In previous fusion blanket studies, the tritium transport rate in flowing LiPb was treated by convection a priori; i.e., the dispersion effect was negligible. In contrast, Taylor dispersion theory shows conflicting results, exhibiting axial transport enhancement via convective flow. In the current paper, the experimental setup consists of a deuterium dissolving tube that substitutes for tritium breeding and a deuterium concentration monitor by LiPb droplets in a vacuum with four nozzles of ϕ = 1.0 mm. The released deuterium mass flux from the droplets was measured using a quadrupole mass spectrometer. An electromagnetic pump circulated 49 L of LiPb at 350°C at a rate between 0.15 and 0.3 L∙min–1 with the corresponding Re number between 600 and 1000, i.e., in the laminar flow range. The dispersion coefficient was analyzed by measuring the temporal distortion of the deuterium concentration profile. The obtained axial dispersion coefficients of dissolved deuterium in LiPb were between 4.6 × 10–2 and 1.2 × 10–1 (m2∙s–1) and approximately seven orders of magnitude greater than those under static conditions. The results agreed with the Taylor dispersion theory, which studied the mass transport enhancement by convection. The applicability of Taylor’s theory to the deuterium flow in liquid LiPb is suggested, whereas the Prandtl number was three orders of magnitude lower and the Schmidt number was one order of magnitude higher than that of the water.

Chemical compatibility of high entropy alloys with liquid PbLi

The chemical compatibility and corrosion resistance of three different high entropy alloys of CrFeV-X, where X = CuW, CuMo, TaW, to stagnant eutectic PbLi was studied. Three nominal HEAs compositions, Cr35Fe35V20W5Ta5 (H-TaW), Cu5Cr35Fe35V20W5 (H-CuW) and Cu5Cr35Fe35V20Mo5 (H-CuMo) were prepared by vacuum arc melting under a low pressure He atmosphere using a non-consumable W electrode and starting from pure elemental pieces with purity > 99.97. Planoparallel specimens of the HEAs were mirror polished and subjected to stagnant PbLi at the eutectic composition, (14.9 ± 1.99) % of Li, at a temperature of 550 °C, corresponding to the working temperature of the WCLL, for increasing times of 168 and 1344 h. After the tests, the microstructure of the samples was cleaned to remove adhered PbLi and analyzed by SEM/EDX and EBSD. SIMS measurements were performed to study the diffusion of into the matrix. From the concentration curves of lithium in the alloys, the diffusion differences of lithium were determined and correlated with the differences in the alloying elements. Furthermore, the corrosion behavior of the alloy showed important differences with the composition of the HEAs and their differences in microstructure. The most significant result is that W has a detrimental effect on the corrosion resistance of PbLi.

Overview of tritium management in WCLL test blanket system of ITER

A key aspect for the achievement of the fuel self-sufficiency of future fusion power plants is the management of tritium in the different stages of the reactor fuel cycle. In particular, one of the greatest challenges for the success of liquid metal breeder concepts such as the Water-Cooled Lithium-Lead (WCLL) is the extraction of tritium from the liquid LiPb eutectic alloy (15.7 at.%Li), which fulfils the main functions of tritium breeder, tritium carrier and neutron multiplier, and the processing in a form suitable for an accurate tritium accountancy. Within this frame, significant advancements have been carried out in the last years as regards the WCLL Test Blanket System (TBS) of ITER. Tritium is mainly produced inside the LiPb and extracted through the Tritium Extraction Unit (TEU). The reference technology for the TEU of the WCLL-TBS is the Gas-Liquid Contactor (GLC) in the packed-column configuration. Tritium is then concentrated in ZrCo getter beds before being sent to the accountancy system where different technologies for both static and dynamic accountancy are adopted in order to provide precise and reliable tritium measurements, which are an essential need in view of an application to the European DEMO reactor. In this paper, the present status and the design solutions foreseen for tritium management in WCLL-TBS are presented and discussed. The implementation and optimization of the conceptual design of the lithium-lead loop (LLP), the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS) have been performed in order to manage tritium concentration in LiPb and in the ancillary systems.

Ab initio study of longitudinal and transverse dynamics, including fast sound, in molten UO2 and liquid Li–Pb alloys

The disparity between the masses of the two components in a binary liquid system can lead to the appearance of a peculiar phenomenon named “fast sound,” which was identified for the first time in Li4Pb several decades ago and later observed in other Li based alloys. However, the exact characteristics and nature of this phenomenon and the reasons behind its appearance have not totally been identified yet. In this work, we analyze the longitudinal and transverse current correlation functions of UO2, Li4Pb, and Li0.17Pb0.83, as obtained from ab initio molecular dynamics simulations. We find that fast sound appears to occur in the two former systems but not in the latter. Additionally, we discuss some of the properties of the liquid mixtures that may be related to the appearance (or absence) of the phenomenon, such as the composition, the polyhedral structure of the melt, and the type of bonding in the system.

Numerical simulation of tritium extraction from liquid PbLi by gas–liquid contactor

As one key component of the fusion reactor, the blanket is in charge of breeding tritium, shielding neutrons and extracting thermal energy for electricity generation. An efficient tritium extraction unit (TEU) is a necessary guarantee for the tritium self-sufficiency in the fusion reactor. The gas–liquid contactor (GLC) is considered to be one of the most efficient devices for the liquid blanket that uses the lead–lithium (PbLi) as the functional materials. Understanding the characteristics of gas–liquid two-phase flow and mass transfer is beneficial for guiding the design and optimization of contactor. However, it is difficult to carry out the tritium experiments due to the scarcity in nature, the strong permeability, as well as the radioactivity. In this paper, the turbulent two-phase model coupled with dilute mass transfer is adopted to numerically study the process of tritium extraction from the liquid PbLi, and the numerical model is evolving step-by-step. Firstly, based on the models of single bubble and helium purge, the effects of different design parameters on the tritium extraction efficiency of the contactor are further investigated (i.e. mass flow rate of helium, initial tritium concentration and liquid PbLi filling height). The results are in good agreement with previous experimental results, and this validates the feasibility of the numerical model used in this work. Then, the entire model of GLC is numerically analyzed, which illustrates that the tritium extraction efficiency of one single unit is only 20.8%. However, the total efficiency can achieve 95% when there are 13 units working together in series. On this basis, the schematic diagram of tritium extraction system is proposed for the fusion blanket.

Parametrization of embedded-atom method potential for liquid lithium and lead-lithium eutectic alloy

Liquid lead-lithium eutectic remains as a promising candidate for various breeding-blanket designs in future nuclear-fusion technologies. The lack of a generalized theory of interatomic forces in the liquid state is reflected on the wide variety of proposed functional forms to describe interatomic interactions even in simple liquids. Computer simulations facilitate the study of liquid metal properties, due to mathematical and experimental challenges. A classical-MD EAM potential is parametrized using mechanical and non-mechanical (melting-point) properties to minimize the arbitrariness of functional forms, where the employed pair potential stems from the liquid-state theory to avoid the issue of the uniqueness of the potential. Enhanced performance is obtained for liquid density, energy, structure, diffusivity and shear viscosity of Li, and their temperature-dependencies. In a similar manner, reference experimental and ab initio MD data are used to parametrize a functional to describe Pb-Li pairwise interactions in liquid Pb-Li alloy, which is used with the derived EAM of liquid Li and a reference EAM of liquid Pb to investigate properties of liquid Pb-Li alloy. Enhanced transferability characteristics are obtained for low-in-lithium liquid Pb-Li melts, where Coulombic interactions are negligible. In specific, the exhibited behaviour of Li in liquid lead-lithium eutectic is consistent with findings from ab initio MD methods, and drastically different from predictions of previous C-MD studies which suggested a substantial segregation of Li atoms instead of dispersion. It is concluded that the functional form of the pair potential and its uniqueness influence both the pure liquid-metal properties and the validity of the potential transferability in multi-component systems, where a theoretical functional results in enhanced performance in pure and alloyed liquid systems.

Modelling the PbLi flow including tritium transport and permeation with GETTHEM

One of the main challenges to be addressed to achieve a reliable electricity production from the EU DEMO reactor is the realization of a closed fuel cycle, for which a suitable Tritium Extraction and Removal System (TERS) is required. One of the possible technologies identified for the EU DEMO TERS is the Permeator Against Vacuum (PAV): the tritium dissolved in the liquid PbLi flowing within several parallel channels will permeate towards the vacuum pumped on the other side of the channel wall (the membrane).A recently-developed model of the tritium permeation across the membrane in the PAV, involving both transport phenomena in the wall and surface processes, was already used to size the EU DEMO PAV. However, besides the component itself, it is important to properly define the interfaces of the PAV in the TERS, and of the TERS in the entire PbLi and tritium loops. The model of such a complex system is therefore implemented here in the Modelica object-oriented language used by system-level tool GETTHEM, that already includes a model of the PbLi loop. The resulting, lumped-parameter component will be able to capture the thermal-hydraulic behaviour of the PbLi, to model the tritium transport in the fluid and to estimate the tritium permeated flux supplied to the tritium processing. Such a model is tested here on a sub-scale circuit to demonstrate its capability to simulate the operation of the EU DEMO TERS using the GETTHEM code.As the physical parameters of the model are subject to a large uncertainty, an uncertainty propagation analysis is also performed, to have a preliminary quantification of the impact of such uncertainties on the model output and, therefore, on the TERS efficiency, and to drive further investigations of these physical properties. In particular, results show how the uncertainty on the solubility constant of hydrogen in PbLi represents the dominant contribution on the total variance, highlighting the need for a better accuracy of such parameter.

Effect of applied stress on the ductility of RAFM steel during long-term exposure in flowing Pb-17Li

The effect of applied stress on the ductility of reduced activation ferritic martensitic (RAFM) steel during long-term exposure in flowing Pb-17Li in the temperature interval 300–400 ℃ was investigated using tensile tests. The ductility of RAFM steels at a stress of 250 MPa decreases appreciably after exposure in following Pb-17Li at 400 ℃ for 1500 h. And a slight ductility loss occurred for the samples tested at 300 ℃, 350 ℃ and 400 ℃ for 500 h. Coarsening of martensitic laths occurred in RAFM steels owing to the thermal aging with a 250 MPa stress for 1500 h in liquid Pb-17Li. Due to the poor ductility of coarse laths, micro-cracks originated on the surface of necking zone during tensile test, which facilitates the penetration of liquid metal. With sufficient liquid Pb-17Li penetrated into cracks continuously, the surface tension of the newly formed fracture surface is declined by liquid metal in the crack tip. The fracture mode was altered from uniform to unstable crack propagation thus leading to a reduction of ductility. There was no depletion of Fe and Cr or Pb infiltration into the substrate for RAFM steels after exposure. The fracture mechanisms were proposed in terms of supply of liquid metal and micro-cracks caused by coarse laths. Accordingly, the reduction of ductility for RAFM steels after exposure in liquid Pb-17Li was the result of combined effect of applied stress and liquid metal.

Experimental setup for study of mechanical performance of materials in liquid Pb-16Li and first results

Dual Coolant Lithium Lead (DCLL) Breeding Blanket, which is using the eutectic lead-lithium alloy (Pb-16Li) as tritium breeder, neutron moderator and coolant, is being developed in the frame of the EUROfusion project. Flow channel inserts (FCI) are insulating components embedded along the Pb-16Li channels inside the DCLL breeding blanket modules developed to minimize the magnetohydrodynamic pressure drop and heat losses. There is an effort to modify the originally designed steel-ceramic-steel layered structure of FCI by eliminating the steel protective layer leading to simple steel-ceramic structure. In this arrangement, the ceramic is directly exposed to liquid Pb-16Li. Demanding operating conditions of the DCLL blanket, in particular elevated temperatures, may give a rise to thermomechanical stresses leading to component failure. Therefore, the assessment of mechanical behaviour of the material in direct contact with Pb-16Li is of primary interest.To study the mechanical properties of the material exposed to corrosive environment of Pb-16Li at elevated temperatures, a dedicated experimental setup was designed. Three-point flexural test is adopted to evaluate the flexural stress-strain material response. Since the experimental procedure is performed in liquid Pb-16Li at high temperatures with an inert atmosphere, the mechanical test takes place in an isolated testing chamber filled with molten Pb-16Li transferred from the melting chamber prior the mechanical test. A specificity of the presented facility is its capability of material testing of three samples within one testing procedure in Pb-16Li at temperatures up to 650 °C.

Performance of Sc-Y-ODS variant of Eurofer steel in stagnant PbLi at 600°C

This paper represents a part of the research of structural materials for fusion systems. The performance of a conventionally formed Eurofer-97 steel and its Sc-Y-ODS variant developed by our group was assessed in a stagnant liquid PbLi environment. The experiments were performed at 600°C for 500 and 1000 hours, and the subsequent microscopical investigation identified a solution-based attack as the dominant corrosion damage of both materials. The metal loss of the conventional Eurofer-97 steel was 66.2 µm after 1000 h exposure, corresponding to a rate of 580 µm/year. In contrast, the ODS Eurofer variant strengthened with various fine Sc-Y complex oxides showed a 70% lower corrosion damage, mainly owing to formation of a Sc-Y-rich Pb surface layer. To understand the degradation process in more detail, both materials were characterized before and after the PbLi exposure in terms of their microstructures and mechanical behavior including hardness, tensile, impact and static fracture toughness tests.

Numerical Simulation of High-Density Ratio Bubble Motion with interIsoFoam

The breeding blanket is one of the fundamental components of a nuclear fusion reactor and is responsible for the fuel production, generating tritium through neutronic capture reaction between lithium and neutrons. Lithium is a liquid PbLi alloy and the helium formed as reaction by-product can coalesce into bubbles, generating a two-phase mixture with a high-density ratio (ηρ∼O5). These bubbles can accumulate and stagnate within the blanket channels with potentially harmful consequences. In this work, the interIsoFoam solver of OpenFOAM v2012 is used to simulate bubble motion for a two-phase mixture representative of the He-PbLi system to test its potential for future developments in the field of fusion. In a first phase, several traditional benchmarks were carried out, both 2D and 3D, and considering the two variants of the VOF method implemented in the solver, isoAdvector and plicRDF. Subsequently, He bubbles of different diameters rising in liquid PbLi (ηρ=1.2×105) were analysed to investigate different regimes. For a Eötvös number (Eo) greater than 10, it was possible to recreate the axisymmetric, skirted, oscillatory regimes and the peripheral and central breakup regimes. For Eo < 10, non-physical deformations of the interface are observed, probably generated by spurious velocities that have a greater impact on the solution for very small bubbles and rising velocities.

Neutronics analyses of COOL blanket for CFETR

In this study, an advanced liquid blanket based on a liquid PbLi technology is proposed, and it is called a supercritical CO2 (S-CO2) cooled Lithium-Lead (COOL) blanket. As a candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), it aims to achieve high tritium breeding and shielding performance. It uses S-CO2 as the coolant for the structural components and PbLi as the tritium breeder and neutron multiplier. To evaluate the neutronics performances of this blanket, a detailed 22.5° three-dimensional neutronics model of CFETR was built using the cosVMPT conversion tool. To enhance the shielding capability of the blanket, a shielding block was designed and added at the back of the manifold to moderate neutrons. Evaluation results of the neutron damage of toroidal field coils (TFC) components confirmed that the blanket meets shielding requirements. In addition, the main neutronics parameters including neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heat, and irradiation damage, were calculated using MCNP5. The results indicated that TBR was 1.205, and the total nuclear heat was 1191.64 MW under the fusion power of 1.5 GW. The peak NWL was 1.97 MW/m2, and the maximum DPA and gas production were 14.80 DPA/FPY, 132.65 He appm/FPY, and 631.30 H appm/FPY.