Fusion Reactor Blanket Research Articles

Growth of α-Al2O3 layer involving abnormal oxides in FeCrAl alloy tube fabricated by WEDM process and electrical insulating performance in fusion reactor blanket

The tubular specimens of FeCrAl alloy APMT (Fe-22Cr-5Al-3Mo) were fabricated by a wire electrical discharge machining (WEDM) process. The oxidation characteristics of the machined surface were investigated by means of the oxidation tests performed in air atmosphere at the temperatures of 1273 K and 1373 K up to 100 h. The abnormal oxides such as CuAl2O4 and ZnAl2O4 were grew on the recast layer which was locally remained on the machined surface even after the mechanical polishing procedure. The α-Al2O3 layer involving the abnormal oxides was formed on the machined surface in the oxidation treatment at 1273 K for 10 h. The apparent electrical conductivity of the layer involving the abnormal oxides was larger than that without the abnormal oxides. Thick Al-rich oxide layer was formed under the abnormal oxides when the oxidation treatment was performed at 1373 K. The apparent electrical conductivity of the oxide layer became lower due to the presence of the thick Al-rich oxide. This oxide layer formed at 1373 K could function as an electrical insulating layer to reduce the magnetohydrodynamic pressure drop in liquid breeder blanket systems of magnetic confinement fusion reactors.

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.

Tensile and creep properties of small specimens of reduced-activation ferritic steel F82H, and the correlation to standard specimen data

Background A reduced-activation ferritic/martensitic (RAFM) steel, F82H steel, is the primary candidate structural material for fusion reactor blanket. Small specimen test technique is essential to develop the blanket materials using limited irradiation volume in high flux neutron field. An international collaboration activity “Towards the Standardization of Small Specimen Test Techniques for Fusion Applications” has been initiated under the framework of the International Atomic Energy Agency Coordinated Research Project for Phase I from 2017 to 2021, and Phase II from 2022 to 2026. The present paper reports the preliminary results on tensile and creep tests as a summary of the above Phase I activity. Methods Tensile and creep tests were conducted at 550 and 650°C, using flat-plate SSJ type small specimens with various gauge thickness ranged from 0.14 to 1.2 mm, while gauge length and width are 5 and 1.2 mm, respectively. In addition, round bar type standard specimens with a gauge geometry of 6 mm in diameter and 30 mm in length was also tested for comparison. Results Tensile yield stress, ultimate tensile strength and uniform elongation were independent of the gauge thickness of SSJ specimens, and agreed with the data from the standard size specimens. On the other hand, total elongation was decreased with decreasing the thickness. In creep tests, rupture time was decreased with decreasing the gauge thickness of SSJ specimens. Standard size specimens exhibited shorter rupture time than the SSJ specimens. Conclusions The SSJ type specimens provided similar tensile parameters to those from the standard specimen, except total elongation. Creep rupture time of the SSJ specimens were different from the standard specimen, and decreased with decreasing the gauge thickness.

High-temperature corrosion testing of titanium beryllides in the presence of water vapor and oxygen

Beryllium-based intermetallics are promising materials for the blankets of future fusion reactors and have potential applications in other areas of the nuclear industry, such as fission reactor reflectors and space technology. Understanding the high-temperature corrosion behavior of these materials in a noble gas medium containing chemically active impurities is essential for evaluating their suitability and guiding their application.This study investigates the high-temperature corrosion of titanium beryllide Be12Ti in the form of plate and grinded samples, produced by JSC “Ulba Metallurgical Plant” (Ust-Kamenogorsk, Kazakhstan). The corrosion tests were conducted under non-isothermal vapor-gas mixture (Ar + D2O or Ar + H2O flowing atmospheres) purging conditions using thermogravimetric (TG) analysis, differential scanning calorimetry (DSC), and mass-spectrometry of the gas phase.As a result of the corrosion tests, new experimental data on thermal effects have been obtained, describing the corrosion processes of Be12Ti samples across a wide range of temperatures and various heating rates in the presence of water vapor in the purge gas. The dependencies of sample mass change under heating conditions have been determined, and the characterization results of the samples before and after high-temperature corrosion tests are presented.Corrosion of titanium beryllides, both for Be12Ti plate and grinded samples, follows similar mechanisms. At around 500 °C, the mass of the samples begins to increase, and hydrogen isotopes are released. The test results indicate that corrosion of titanium beryllides with varying surface inhomogeneities proceeds similarly within the temperature range of 500–900 °C, showing a linear dependence on temperature.The results revealed significant insights into the oxidation mechanisms and the formation of corrosion products, which are crucial for optimizing the material's performance in fusion reactor environments.

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

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

Lithium lead titanate (Li2PbxTi1-xO3, 0.1

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

Impact of Reverse Buoyancy on Heat and Mass Transfer in MHD Flow of PbLi Within a Poloidal Duct in a Typical DCLL Liquid Breeder Blanket

The typical dual-coolant lead-lithium (PbLi) design of a liquid breeder blanket in a magnetic confinement fusion reactor involves the utilization of PbLi as the working fluid to effectively remove neutron heat. However, the nonuniform heating of neutrons with a significant radial gradient induces a buoyancy effect, resulting in the formation of vortexes ices within the downward flow duct. These vortexes have an adverse impact on the heat and mass transfer characteristics of the magnetohydrodynamic (MHD) flow of PbLi. The simulations in this work employed a MHD buoyant mixed-convection solver to resolve the characteristics of PbLi flow and a one-way coupled Lagrangian method to analyze the qualitative characteristics of tritium transport in PbLi flow. The results indicate that buoyant reverse flow can create vortexes that contain hot spots in the PbLi fluid, which can significantly impede heat transport. Additionally, the vortex causes tritium recirculation in the flow field and retention, resulting in adverse effects on tritium transport.

Simultaneous measurements for fast neutron flux and tritium production rate using pulse shape discrimination and single crystal CVD diamond detector

This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.

Modeling of tritium release behavior of biphasic Li2TiO3–Li4SiO4 ceramics

Biphasic Li2TiO3–Li4SiO4 ceramics have been proposed as advanced tritium breeder for solid-type breeding blankets of fusion reactors. However, the mechanism of tritium release from biphasic Li2TiO3–Li4SiO4 ceramics remains to be elucidated. In this study, an integrated numerical model for tritium release from biphasic Li2TiO3–Li4SiO4 ceramics was constructed for the first time. In the proposed model, besides the mass transfer steps for tritium release from single-phase ceramics (e.g. Li2TiO3 and Li4SiO4), a mass transfer step between the interfacial layers of Li2TiO3 and Li4SiO4 was also included to take into account the effect of two-phase interface in Li2TiO3–Li4SiO4 ceramics. By using this model, numerical simulations of tritium release from Li2TiO3–Li4SiO4 ceramics with different phase ratio and grain size were performed. The simulation results showed that the mass transfer of tritium atoms between the interfacial layers of Li2TiO3 and Li4SiO4 can effectively enhance tritium release from Li4SiO4 in biphasic Li2TiO3–Li4SiO4 ceramics. Furthermore, increasing phase ratio of Li2TiO3 and decreasing grain size of Li2TiO3 were both beneficial to tritium release from biphasic Li2TiO3–Li4SiO4 ceramics at lower temperature. The modeling work consolidated the presumption that the two-phase interface could play an important role in the tritium release process of biphasic Li2TiO3–Li4SiO4 ceramics.

Evaluation of the impact of plasma operation scenarios on the fusion reactor blanket design using an integrated numerical plasma and neutronics analysis suite

A very limited operation scenario has been used for fusion neutronics for the design of tritium breeding blankets. However, many advanced operation scenarios are under development and the impact of choosing a plasma operation scenario for fusion neutronics, which requires integration of neutronics analysis with plasma analysis, needs to be evaluated. An automated process from plasma analysis to neutronics analysis to create a viable fusion plasma neutron source was considered using a few different codes. In this work, we develop an integrated suite combining plasma modeling codes and neutronics codes. McCARD, a Monte-Carlo neutron-photon transport simulation code, is integrated into TRIASSIC, a tokamak reactor integrated automation suite for simulation and calculation, as the module for neutronics analysis of various plasma operation scenarios. McCARD integration allows TRIASSIC to perform neutronics analysis along with analytical and predictive plasma simulations. To evaluate this integrated suite, KSTAR experimental data were used to calculate the neutron flux at the first wall via TRIASSIC with the McCARD module in three different cases. First, the effect of plasma shape parameters, elongation and triangularity, on neutron flux is analyzed. Second, plasma scenario with the changes to the shape and the performance of plasma over time is studied to evaluate the time varying simulation. Third, two different plasma scenarios are compared to analyze the impact of plasma performance on neutron flux.

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

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

Numerical and experimental analysis of magneto-convective flows around pipes

Liquid metal flows exposed to intense magnetic fields play a fundamental role in the development of blankets for nuclear fusion reactors, where they serve to produce the plasma-fuel component tritium and transport the generated heat. When the liquid metal circulates in the blanket, it interacts with the plasma-confining magnetic field leading to the induction of electric currents. Electromagnetic forces and thermal buoyancy affect significantly velocity and pressure distributions. The resulting magneto-convective flow has peculiar features that have to be taken into account in order to evaluate heat transfer properties in the liquid metal. In the breeding zone of the water-cooled lead lithium blanket concept, the volumetric heat released in the liquid metal is removed by water-cooled circular pipes that represent obstacles for the liquid metal flow. In order to improve the understanding of the underlying physical phenomena and obtain a database for code validation, model experiments have been performed to investigate magneto-convective flow and heat transfer at two differentially heated parallel horizontal tubes immersed in a box filled with liquid metal. Among other properties, electric potential has been recorded on the surface of the test-section and compared with 3D numerical simulations. Computational results provide the basis for interpretation of the measured data, since they allow differentiating between flow-induced electric potential and thermoelectric effects.

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.

Study of hydrogen sorption and desorption processes of zirconium beryllide ZrBe2

One of the promising intermetallic compounds for use in nuclear and fusion reactors, as well as in hydrogen energy technology, are intermetallic compounds of beryllium with metals such as Ti, V, Zr and Nb. Beryllium-based intermetallics are not only a promising material for blankets of future fusion reactors, but can also be utilized in other areas of the nuclear industry, including fission reactor reflectors, rocket and space technology. The interest in studies of the interaction of hydrogen isotopes with beryllides is related to the accumulation of tritium and helium in the material, formed as a result of nuclear reactions under neutron irradiation, and is caused by the need for understanding the processes arising from such interaction.In this work, hydrogen sorption and desorption processes of zirconium beryllide ZrBe2 produced by industrial technologies at JSC “Ulba Metallurgical Plant” (JSC UMP) were investigated. The experiments were performed by the Sievert’s and thermal desorption spectrometry (TDS) methods. In TDS experiments deuterium was chosen for samples saturation to reduce the possible error in determining fluxes associated with the release of hydrogen from the elements of the vacuum chamber.An equation for hydrogen solubility in zirconium beryllide ZrBe2 was obtained by processing the experimental data obtained by the Sievert’s method:S=0.46±0.1molm3∙Pa12∙exp-13.3±0.5kJmolRTThe temperature intervals of formation and decomposition of two different hydride phases (deuterides) of zirconium beryllide ZrBe2 at ∼ 600 K and ∼ 900 K have been established in TDS experiments.

Non-modal stability analysis of magneto-hydrodynamic flow in a single pipe

A complete understanding of the stability of fluid flows under varying magnetic field profiles is imperative for achieving control of plasma and operating fluids in the blankets of future fusion reactors. In this context, the primary objective of this study is to investigate the influence of varying magnetic profiles on the flow regime of a generic fluid, which is representative of both thermonuclear plasma and conductive fluids within a nuclear fusion reactor. To this aim in this work non-modal stability theory is adopted to perform stability analysis of a magneto-hydrodynamic (MHD) flow in an infinite circular pipe in order to study the effects of the magnetic field on the fluid dynamics of the pipe flow. In particular, the effects on the general stability of two magnetic field profiles are compared with the reference case of a pipe Poiseuille flow without magnetic field. Firstly, the classic modal stability technique is employed to study asymptotical stability. Then, non-modal stability analysis is applied to magneto-hydrodynamic pipe flow to study the system's response for a finite time immediately after a perturbation. Fourier–Chebyshev Petrov–Galerkin spectral method is used to compute the eigenvalues and pseudospectra of the weak formulation associated with the linearised system. Investigations on the dependence of spectra and transient growths on the specific magnetic profiles are conducted for different values of perturbation wave numbers. The obtained results show that in general the magnetic field has an effect of stabilization on the system, which depends on the specific magnetic profile considered. In addition, the non-modal stability analysis reveals that the inclusion of the magnetic field mitigates the effects of perturbations also in the short term, a phenomenon that cannot be seen using only modal stability analysis.

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.

Microfabrication of Li4SiO4 pebbles with complex pore structure via digital light processing: Accuracy control and properties optimizing

Li4SiO4 ceramic pebble is deemed as the most attractive tritium breeder for fusion reactor blanket. In this work, we fabricated a novel Li4SiO4 pebble with micrometer-scale pore structure by digital light processing to verify the feasibility of complex pore structure manufacturing for future breeders. The effects of processing parameters on printing accuracy, microstructure and mechanical properties of Li4SiO4 pebbles were investigated. For synthesizing the pure Li4SiO4 phase, the optimal stoichiometric ratio of Li2SiO3 to Li2CO3 was 1:1.09 in the premixed powder. The photosensitivity of dual-phase ceramic suspension was investigated according to an attenuation model, and the overgrowth was reduced to ∼50 μm by using the dimension formula. After optimizing the post-processing, the radial pore structure Li4SiO4 pebbles with excellent crush load (50.3 N, standard deviation = 4.4 N) and suitable open porosity (18.3 %) can be yielded. This work provides a fresh solution to control the pore structure of tritium breeding pebbles in an unprecedented micrometer scale.

Calculation of Temperature Fields in a Lithium Ceramic Pebble Bed during Reactor Irradiation in a Vacuum.

Two-phase lithium ceramic Li2TiO3-Li4SiO4 is considered as a tritium multiplier for use in the solid blanket of fusion reactors. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials under neutron irradiation can only be obtained from experiments in fission research reactors. At that, irradiations in vacuum give the possibility to register even very fast gas release processes (bursts) from the ceramics' voids and pores, although it reduces the thermal conductivity of the pebble bed. The purpose of this work was to simulate the heating of mono-sized pebble bed (1 mm in diameter) of two-phase lithium ceramic 25 mol%Li2TiO3+75 mol%Li4SiO4 in an ampoule device during neutron irradiation at the WWR-K research reactor under vacuum conditions, and to determine experimental parameters in order to prevent heating of the lithium ceramics up to the Li4SiO4-Li2SiO3 phase transition temperatures (>900 °C). For the first time, it was obtained that the effective thermal conductivity of a 1 mm mono-sized pebble bed of 25 mol%Li2TiO3+75 mol%Li4SiO4 significantly decreases (four times) when it is irradiated with neutrons in a vacuum (at a helium pressure of approximately 10 Pa), compared to a similar calculation at 100 kPa of helium (when the He sweep is used). It was concluded that it is difficult to evaluate the maximal temperature of the ceramics in the capsule by measuring the temperature of its outer metal wall (according to thermocouple readings) without using the results of thermophysical calculations for each type of ceramic, taking into account its quantity, specific heat release and pebble size(s). To control the temperature of the ceramics during an irradiation experiment in a vacuum, an in-capsule thermocouple should be used, placed in the center of the pebble bed. Measuring the temperature of the pebble bed based on the capsule wall temperature can lead to overheating of the ceramics and phase changes.

Numerical Investigation of the Effect of the Wall Properties on Downward Mercury Flow and Temperature Fluctuations in a Vertical Heated Pipe under a Transverse Magnetic Field

The work is motivated by the need to study the effect of abnormal quasi-periodic high-amplitude temperature fluctuations in a liquid metal flow in pipes and channels, which arise under conditions close to those specific for liquid metal (PbLi alloy, etc.) hydrodynamics and heat transfer in blankets of tokamak fusion reactors. Such temperature fluctuations in the cooling channels of the fusion reactor blanket pose a threat to the strength of the channel walls. The influence of the physical properties of the channel walls and potential fouling of their inner surface on the generation of these fluctuations has not been studied as yet. Therefore, this problem has been numerically investigated for a MHD mixed convection of a liquid metal in a round tube for a downward flow of mercury with the one-side heating condition under a strong transverse magnetic field using conjugate problem statement by the LES (Large Eddy Simulation) method. Numerical simulation was performed for four cases: (i) the effect of thermophysical and electrophysical properties of the steel pipe wall is neglected, (ii) the effect of wall material properties is included, (iii) the effect of the moderate contact electrical resistance caused by a layer of fouling deposits or oxides on the inner surface of the pipe is accounted for, and (iv) the resistance of the thin fouling layer is assumed to be very high. The predictions for the cases (i) and (iv) are in good agreement with the experimental data and the results of the direct numerical simulation. They demonstrate the existence of quasi-periodic abnormal high-amplitude temperature fluctuations in the fluid with a frequency of approximately 0.14 Hz. With a relatively low electrical resistance of the fouling film (case iii), the frequency of high-amplitude temperature fluctuations was considerably lower (0.07 Hz). In the absence of electrical contact resistance on the inner surface of the steel pipe (case ii), high-amplitude velocity and temperature fluctuations were not revealed in the fluid. Thus, it was shown for the first time that the physical properties of a wall and the electrical resistance between the fluid and the electrically conducting wall were responsible for the development of abnormal temperature fluctuations in the liquid and the wall and control of their amplitude and frequency. The causes of the revealed effects are discussed.

Depth-resolved and time-resolved investigations of Er2O3 ceramics after static lithium corrosion by using LIBS at different pressures

Liquid metal lithium can cause severe corrosion on the surface of fusion reactor blankets and first-wall metal materials. Oxidized erbium ceramic has been widely studied as a candidate material for the self-cooling blanket system of tokamak fusion reactors. In this study, time-resolved picosecond laser-induced breakdown spectroscopy (ps-LIBS) was used to investigate the time evolution process of laser-induced plasma (LIP) of Er2O3 samples in air at different pressures. The pressure-time windows with high signal-to-background ratio and high signal-to-noise ratio under LTE conditions were determined. The composition depth distribution of the Er2O3 ceramic after static corrosion by lithium was analyzed. The number of laser pulses were linearly proportional to the laser ablation depth as measured by a three-dimensional profilometer. As the number of pulses increased, the relative intensities of the 610.353 nm and 812.645 nm characteristic lines of the corroding element Li gradually decreased, while the relative intensities of the 405.547 nm and 405.951 nm characteristic lines of the base element Er gradually increased. The laser pulse number at which the relative peak intensities of Er and Li reached a steady state showed a trend of first increasing and then decreasing with increasing pressure. In addition, the results of laser ablation depth under different environmental pressures indicated that the depth resolution of the analysis of anti-corrosion Er2O3 ceramic materials can be controlled by the environmental gas pressure.