Reactor Materials Research Articles

Mechanism of dynamic recrystallization of a FeCrAl alloy during hot compression

FeCrAl alloys have the opportunity to replace current Zr-based cladding materials in light water reactors, and dynamic recrystallization (DRX) plays an important role on microstructural controlling of FeCrAl alloys during hot or warm deformation, such as forging, hot rolling, extrusion. In this research, dynamic recrystallization mechanism of FeCrAl alloy has been systematically investigated by microstructure characterization after isothermal compression. Results show that deformation strain rate and temperature could significantly influence microstructure and DRX mechanism of FeCrAl alloy. When deformed at 0.1s−1, prior grains become long strip with serrated boundaries and are refined by geometric dynamic recrystallization (GDRX). Prior grains are not separated by high angle grain boundaries (HAGB) even deformed at 1200 °C, because strain is too small to impingement of HAGB. Whereas, when deformed at 10s−1 strain rate, there are lots of sub-boundaries with about 2° misorientation within grains, and several strain-free grains are found located at boundaries when deformed at 900 °C or 1000 °C, while the prior grains are divided by HAGB and equiaxed grains are formed at 1100 °C. It was found that mechanism of DRX changes from discontinuous dynamic recrystallization (DDRX) to continuous dynamic recrystallization (CDRX) with increasing temperature from 900 °C to 1100 °C. The phenomenon results from the dislocation movement behaviors changes with increasing temperature. This research will provide meaningful guidance for industrial application of FeCrAl alloys.

Determination of the Surface-Rate Coefficients of Deuterium Through Nickel by Gas-Driven Permeation Experiments in Surface-Limited Regime

Tritium permeation through structural materials of fusion reactors is an important issue from both safety and tritium self-sufficiency points of view. Mutual influences among hydrogen isotopes are also important in assessments of tritium permeation fluxes, and although some theoretical and experimental works have been carried out so far—which have all focused on steady-state permeation—their conclusions are insufficiently clear about how one isotope affects the permeation of another. This was motivation to further investigate this problem, both theoretical and experimental in a non-steady-state approach and also in the surface-limited regime (SLR) of permeation. After initial theoretical work, a dedicated experimental installation for gas-driven permeation experiments was designed and assembled. Then, initial experimental work was completed for testing monoisotope permeation of deuterium through a very thin (0.075-mm) nickel membrane in the temperature range of 473 to 773 K and for a driving pressure of deuterium gas in the range of 10−2 to 1 Pa. This work was done in preparation for subsequent multi-isotope experiments, with the proposed goal to observe how hydrogen affects the permeation of deuterium. The main objectives of the work were to confirm experimentally the achievement of the SLR of permeation for the selected conditions of testing; to determine experimentally the kinetic coefficients of surface transport for deuterium in nickel; and to test and validate the experimental setup, procedures, and methods used and their reliability to more stringent requirements of the multi-isotope experiments. Within this paper, the experimental setup and all the operating procedures used both in calibration operations and in permeation experiments are presented in detail, as well as their results. The obtained results confirmed that permeation occurred in the SLR in the tested range of pressure for each testing temperature. The surface-rate coefficients of dissociation and recombination were both determined experimentally. The values obtained for the dissociation coefficient were in very good agreement with other similar experimental data available in the literature. For the recombination coefficient, agreement was not quite satisfactory, but comparison could be made with values calculated based only on the dissociation and solubility coefficients, as data for these coefficients for the nickel-hydrogen system (determined directly from gas-driven permeation experiments) were not found in the literature. However, these results indicated good reliability of the installation, its calibration, and operating procedures in order to proceed to experimental testing of multi-isotope permeation.

The behavior of helium bubble evolution under neutron irradiation in different tungsten surfaces

Tungsten, as a kind of material for the plasma-facing wall in fusion reactor and tokamak, is subjected to irradiation by the high-energy neutron and the high fluxes of hydrogen and helium plasmas. Especially, in the presence of helium (He) bubbles, neutron irradiation can significantly exacerbate the deterioration of the material's mechanical properties. However, due to experimental limitations, the evolution of He bubbles in tungsten irradiated by neutrons and the underlying mechanisms remain unclear. Here, molecular dynamics (MD) simulation is used to investigate the behavior of He bubbles in tungsten with different crystal surfaces irradiated by neutrons. We observed the complete evolutionary process of the He bubble, i.e. standing, expanding, and bursting. During this process, high-energy neutron transfers energy to atoms within the He bubble through cascade collision, leading to a rapid increase in energy and pressure of the He bubble, and then the rapid expansion of the He bubble makes the pressure rapidly released, ultimately leading to the rupture of the helium bubble. Furthermore, we also explore the evolution of the He bubble under varying conditions (such as temperature, neutron energy, and He/vacancy ratio) within the same crystal surface. We found that at lower temperatures and neutron energy with a higher He/vacancy ratio, it tends to form pinhole-like channels to release helium atoms, resulting in the He bubble bursting. On the contrary, it tends to form a hill-like expansion to spurt helium atoms out, leading to the bubble bursting. Additionally, it was found that the direction and time of He bubble rupture varied under different crystal surfaces. He bubbles under the (112) surface burst the slowest with excellent corrosion resistance. These findings provide new insights into the behavior of irradiation resistance of fusion reactor materials under extreme conditions and offer significant guidance for the improvement and selection of optimal materials for plasma-facing walls.

Chemical composition based machine learning model to predict defect formation in additive manufacturing

With a goal of exploiting additive manufacturing to improve the manufacturing of existing reactor materials, we developed a chemical composition-based machine learning model to predict the printability of any given alloy in laser powder bed fusion (L-PBF) using experimental data from peer-reviewed literature. We defined printability as the ability to avoid defects like cracking, balling, porosity, and lack of fusion, that are caused by thermal stresses (during solidification or liquation), molten pool disintegration into disconnected small beads or lack of heat input respectively. Our models predict the tendency of balling defect formation and porosity percentage for a given composition, under a given set of processing conditions. To predict the likelihood of balling defect, three models: a random forest classifier, a gradient boost regressor and a neural network were trained on a dataset containing both single and multi principal element alloys. The neural network model showed the highest accuracy of 92.3 % in predicting the balling defect formation. A random forest regressor, gradient boost regressor and neural network were trained and tested on a dataset of various alloys to predict porosity. The random forest regressor showed the best predictions with an R2 score of 0.97. The models also revealed the relative importance of the input descriptors on defect-formation tendency. Of particular significance was the identification of carbon as an important element in determining the occurrence of balling and percent porosity in alloys like steel, as well as being moderately important to the percentage porosity in other alloys as well as steel. Manganese was also identified as a key descriptor for the percentage of porosity in steel and other alloys. Manganese's low thermal conductivity and consistent presence in the dataset is the likely cause for its contribution. Carbon's role is attributable to its relatively high specific heat and high melting temperature. Our model serves as a swift, chemistry-based tool to design experiments and find modified compositions better suited for additive manufacturing.

Cross sections for 14 MeV neutron interaction with lutetium isotopes and their theoretical excitation functions* *Supported by the Natural Science Foundation of Henan (232300420130) and National Natural Science Foundation of China (11575090)

The cross-sections for the 175Lu(n,p)175Yb, 175Lu(n,α)172Tm, 176Lu(n,α)173Tm, 175Lu(n,2n)174mLu, and 175Lu(n,2n)174gLu reactions at 13.57, 14.03 14.62, and 14.86 MeV neutron energies were measured using an activation technique. The theoretical excitation functions of these reactions were calculated using the Talys-1.95 code. The reaction cross-section data experimentally obtained were analyzed and compared with experimental data reported in the literature, data from five major evaluated nuclear data libraries of IAEA, and theoretical values based on Talys-1.95. The data obtained at some neutron energies agree with some of the data reported in the literature and theoretical values based on Talys-1.95. The consistency of the theoretical curves of excitation functions based on Talys-1.95 with the data obtained in this study and those reported in the literature is higher than that of the evaluation curves of excitation functions for the 175Lu(n,p)175Yb, 175Lu(n,α)172 Tm, and 176Lu(n,α)173Tm reactions. This study is helpful because it provides new evaluated reaction cross-section data on lutetium (which is a fusion reactor material), improves the quality of neutron-induced reaction cross section data libraries, and advances the research on related applications.

Cavity Swelling of 15-15Ti Steel at High Doses by Ion Irradiation.

The titanium-stabilized austenitic stainless steel Fe-15Cr-15Ni, which shows enhanced resistance to irradiation swelling compared with more traditional 316Ti, has been selected as a core material for fast reactors. Data on the evolution of irradiation swelling in 15-15Ti steels at very high doses, which cannot be easily achieved by neutron irradiation, are still lacking. In this paper, the swelling behavior of the titanium-modified austenitic stainless steel 15-15Ti was investigated by pre-implantation of He at room temperature followed by Ni-ion irradiation at 580 °C to peak doses of 120, 240 and 400 dpa. Relatively small cavities were observed in the zone of helium implantation, while large cavities appeared in the region near the damage peak. A correction formula for the dpa curve was proposed and applied to samples with large swelling. It was found that the steady-state swelling rate of 15-15Ti remains at ~1%/dpa even at high doses. By comparing the swelling data of the helium-implanted and helium-free regions at same doses, 70 dpa and 122 dpa, the suppression of swelling by excessive helium can be deduced at such doses.

Study on photoluminescence properties of Er2O3 materials as irradiation damage and temperature sensors

Photoluminescence (PL) properties of Er2O3 specimens were examined by using visible lasers (532 nm and 635 nm) and a UV LED light source (365 nm) to investigate the applicability for irradiation damage monitoring of materials in fusion reactors. Both in the laser induced and UV light induced PL spectra, green (510–590 nm) and red (630–725 nm) luminescence was observed. In the spectrum measurements on specimens with different crystallinities, it was confirmed that an intensity of the red luminescence weakened significantly compared with that of the green luminescence in an Er2O3 specimen with a lower crystallinity. The results indicate that the PL measurements of Er2O3 materials could be applicable for the irradiation damage monitoring in fusion reactors. The luminescence property of ion beam irradiated Er2O3 showed that information of irradiation damages could be kept up to ∼ 300 ℃ and almost recovered at 700 ℃. Based on the obtained luminescence properties, positions in a fusion reactor where Er2O3 materials could be used as irradiation damage sensors are proposed. Changes in PL spectra at high temperatures up to ∼ 400 ℃ indicate the possibility that the Er2O3 materials might be applicable also for temperature monitoring of in-vessel components during reactor maintenance periods.

Understanding the Subsurface Microstructure and Thermal Behavior of Model Oxide Dispersion Strengthened Alloys Through FIB_SEM and TEM

AbstractA potential but an underexplored application of FIB_SEM is its ability to image the subsurface microstructure and capability for an associated chemical analysis. In this article, agglomerated microstructures of two model oxide dispersion strengthened alloy systems, consisting of ZrO2 and Y2O3 dispersions, are evaluated to understand its elevated temperature behaviors. The systems under evaluation are relevant as candidate nuclear structural materials for next-generation fast breeder reactors, and FIB_SEM technique is effectively used along with TEM and EDS for a comprehensive understanding of the material microstructure, and the results are discussed. Distinct microstructures are observed for the two systems with a difference in crystallite size distribution and presence of micron-sized dispersoids in Y2O3. The varied behavior of dispersoids is discussed in terms of its pre-annealing microstructures, and the superiority of ZrO2 over Y2O3 as a dispersoid for ODS alloys is emphasized.

High-temperature interaction of water vapor with lithium ceramics Li2TiO3

Selection of materials for fusion reactors and, in particular, for blanket systems is one of the most important tasks that determine the development of fusion energy. Lithium ceramics Li2TiO3 is considered as prospective candidate for solid breeders of future fusion reactors' blankets. The paper presents the results of high-temperature tests of Li2TiO3 + 5 mol% TiO2 ceramics, performed under conditions of blowing with argon containing impurities of water vapor of different isotopic compositions (H2O, HDO, D2O) in the range of temperature from 100 °C to 1400 °C. The vapor–gas mixture for the experiment was prepared in a special mixing tank. The interaction of Li2TiO3 + 5 mol% TiO2 ceramics with chemically active gases and water vapors was investigated by thermogravimetry (TG), differential scanning calorimetry (DSC) and mass spectrometry (MS) methods. As a result of experiment, the dependences of the sample mass change during heating under single purge gas composition was obtained. Mass spectra that characterize the changes in the qualitative composition of the gas mixture in the reaction chamber with the investigated sample were recorded and subsequently analyzed. An isotope effect is observed during the experiment: the release of D2 gas begins significantly earlier than the release of HD gas. The observed effect appears only in the high-temperature region (above 1050 °C), which characterizes the investigated material as almost completely chemically neutral towards water vapor.

Design and safety features of SALUS-100: A long fuel-cycled Sodium-cooled fast reactor

The paper introduces the innovative development of an SFR-based Small Modular Reactor (SMR) known as SALUS-100, with a 100 MWe electrical output, developed by KAERI. This innovative design, derived from the previous TRU transmutation burner SFR concept, represents a significant step forward in establishing a new model for SMRs and offers diverse applications, including on– and off-grid electricity generation, district heating, and industrial heat supply. The reactor features of SALUS-100 incorporate an integrated pool-type primary system configuration and a long fuel-cycle core concept with a metallic alloy nuclear fuel. Its design concept prioritizes preventing severe accidents by capitalizing on the inherent safety and reliability features achieved through emphasizing passive design characteristics. The paper describes the engineering design and safety analyses of SALUS-100, highlighting advanced design concepts in reactor core, fuel, materials, and systems. Additionally, the key research and development endeavors, focusing on vital areas such as enhancing system design capability, validating designs, demonstrating safety, and innovating in metal fuel technology, are briefly discussed, emphasizing the potential of SALUS-100 in revolutionizing the SMR market and utilizing efficient nuclear energy in the future.

Microstructural changes in He-irradiated V-Cr-Ti alloys with low Ti addition at 700 °C

V-Cr-Ti alloys with reduced Ti contents are expected to be good candidates as recyclable structural materials for fusion reactors. Herein, He irradiation was applied to V-4Cr-xTi (x = 0 to 4) alloys to investigate the additional effect of Ti and gas interstitial impurities on their microstructural evolution and irradiation hardening. Following He irradiation of the specimens at 700 °C with a maximum damage depth of 0.5 dpa, transmission electron microscopy and nanoindentation hardness tests were conducted. The microstructural evolution of the titanium-oxycarbonitride (Ti(CON)) precipitates formed during He ion irradiation was observed. The behavior of irradiation hardening is explained by the formation process of Ti(CON) precipitates and the magnitude of irradiation hardening depends on the amount of gas interstitial impurities that contribute to precipitate formation.

Navigating Pyrolysis Implementation-A Tutorial Review on Consideration Factors and Thermochemical Operating Methods for Biomass Conversion.

Pyrolysis and related thermal conversion processes have shown increased research momentum in recent decades. Understanding the underlying thermal conversion process principles alongside the associated/exhibited operational challenges that are specific to biomass types is crucial for beginners in this research area. From an extensive literature search, the authors are convinced that a tutorial review that guides beginners particularly towards pyrolysis implementation, from different biomasses to the thermal conversion process and conditions, is scarce. An effective understanding of pre-to-main pyrolysis stages, alongside corresponding standard methodologies, would help beginners discuss anticipated results. To support the existing information, therefore, this review sought to seek how to navigate pyrolysis implementation, specifically considering factors and thermochemical operating methods for biomass conversion, drawing the ideas from: (a) the evolving nature of the thermal conversion process; (b) the potential inter-relatedness between individual components affecting pyrolysis-based research; (c) pre- to post-pyrolysis' engagement strategies; (d) potential feedstock employed in the thermal conversion processes; (e) the major pre-treatment strategies applied to feedstocks; (f) system performance considerations between pyrolysis reactors; and (g) differentiating between the reactor and operation parameters involved in the thermal conversion processes. Moreover, pre-pyrolysis activity tackles biomass selection/analytical measurements, whereas the main pyrolysis activity tackles treatment methods, reactor types, operating processes, and the eventual product output. Other areas that need beginners' attention include high-pressure process reactor design strategies and material types that have a greater potential for biomass.

Irradiation Hardening and Microstructure Study of MAX-Phase-Dispersion-Strengthened Vanadium Alloy under Self-Ion Irradiation

V-4Cr-4Ti alloy is one of the candidate structural materials for future fusion reactors due to its desirable characteristics. In our previous research, MAX-phase-dispersion-strengthened vanadium alloy (V-4Cr-4Ti-1.5Y-0.3Ti3SiC2), prepared through mechanical alloying, showed excellent thermal stability and creep resistance and was expected to have good radiation resistance. This study investigates the effects of 2.5 MeV V2+ ion irradiation on V-4Cr-4Ti-1.5Y-0.3Ti3SiC2 and V-4Cr-4Ti alloys at 500 °C, with peak damage of 0.8, 3.5, and 6.1 dpa. Transmission electron microscopy and nanoindentation were used to examine the changes in microstructure and hardness before and after irradiation. The microscopic analysis reveals that dispersed nanoparticles maintained good stability under irradiation. Defect clusters grow with increasing irradiation doses in both materials. The nanoindentation results show that V-4Cr-4Ti-1.5Y-0.3Ti3SiC2 has higher initial hardness and lower irradiation hardening, indicating better resistance to radiation hardening than V-4Cr-4Ti. This research serves as a valuable reference for the assessment of the irradiation resistance of Ti3SiC2-dispersion-strengthened V-4Cr-4Ti alloy.

Assessing the master curve reference temperature T0 estimated from the Charpy impact transition curve and consequences on the correlation between Charpy DBTT- and fracture toughness T0-shifts

Integrity assessment of reactor pressure vessels of light water reactors follows well established procedures relying on the surveillance program and defined by codes, standards and nuclear regulatory requirements. These procedures were developed in times where scientific knowledge was limited increasing therefore conservative safety margins to overcome the scientific shortcomings.Taking profit from the development of a number of test techniques and evaluation, it is possible nowadays to propose an alternative interpretation of the test results that provides a complementary information to the standard regulatory approach. In this work, the focus is on the Charpy impact test as it plays a critical role in surveillance programs monitoring reactor vessel degradation induced by neutron irradiation. The development of methodologies allowing experimental determination of fracture toughness from small size specimens triggered alternative approaches of extracting fracture toughness from the Charpy impact test. One of those approaches was recently successfully developed using the instrumented Charpy impact test from which the dynamic and static fracture toughness can be estimated with a reasonably good accuracy. This paper aims to provide an equally performant but simplified procedure allowing estimation of the master curve reference temperature, T0, from the Charpy impact transition curve applied to a large variety of reactor materials and irradiated conditions.

Revealing the impact of sink strength, injected interstitial and grain growth on the temperature-dependent depth distribution of voids in nanocrystalline Ni

The application of electrodeposited (ED) Ni coating on structural materials of molten salt reactors has emerged as an attractive option for improving the corrosion resistance of the base alloy. However, the radiation response of the Ni coating, especially the void swelling, is an important aspect to be considered for potential in-pile applications. The study investigates depth-dependent void swelling behavior in 20 nm grain-sized ED nanocrystalline (NC) Ni, irradiated with 1.4 MeV Ni+ ion irradiation at 350–550°C to a peak displacement damage of 18.5 dpa. The injected interstitial effect is absent at 350°C; however, it becomes prominent at 450°C and less at 550°C. This is atypical of ion irradiations in coarse-grained Ni, where the effect becomes more pronounced at lower temperatures. The occurrence of voids at all temperatures goes deeper than TRIM predictions; this effect is more significant at 350°C, surpassing any previous findings in coarse-grain systems. Depending on the irradiation temperature, various types of void distributions about grain boundaries are observed: grain boundaries with no void denudations or one-sided and double-sided denudations. Here, we demonstrate that the instability of NC grains due to the combined influence of temperature and irradiation and the high fractions of internal sinks in NC Ni are responsible for void distributions that may be unique to NC systems.

Enhanced thermal-mechanical properties of rolled tungsten bulk material reinforced by in situ nanosized Y–Zr–O particles

Tungsten is the most promising plasma facing material for fusion reactors. Rolled W–Y2(Zr)O3 bulk material has been successfully produced in this study for future fusion engineering applications. The introduction of Zr is conducive to the refinement of the second phase particles. Nano-sized Y–Zr–O particles are observed in the powder and bulk samples. Related results show that the Y–Zr–O particle dispersion distribution improves the heat load resistance of W–Y2(Zr)O3 composite material. For four-point bend experiments in the same sampling direction, the DBTT of W–Y2(Zr)O3 composite materials is lower compared to the pure tungsten. For the same material, the DBTT of the material was selected for testing along the RD direction is lower compared to the material was selected for testing along the TD direction. Findings of this study provide suggestions for the subsequent industrial preparation of nanoscale particle-doped tungsten materials.

Recent progress of particle electrode materials in three-dimensional electrode reactor: synthesis strategy and electrocatalytic applications.

With the complete promotion of a green, low-carbon, safe, and efficient economic system as well as energy system, the promotion of clean governance technology in the field of environmental governance becomes increasingly vital. Because of its low energy consumption, great efficiency, and lack of secondary pollutants, three-dimensional (3D) electrode technology is acknowledged as an environmentally beneficial and sustainable way to managing clean surroundings. The particle electrode is an essential feature of the 3D electrode reactor. This study provides an in-depth examination of the most current advancements in 3D electrode technology. The significance of 3D electrode technology is emphasized, with an emphasis on its use in a variety of sectors. Furthermore, the particle electrode synthesis approach and mechanism are summarized, providing vital insights into the actual implementation of this technology. Furthermore, by a metrological examination of the research literature in this sector, the paper expounds on the potential and obstacles in the development and popularization of future technology.

Demonstration of Helide formation for fusion structural materials as natural lattice sinks for helium

Fusion power holds promise as an ultimate energy source. However, achieving true sustainability in fusion energy requires addressing the embrittlement of polycrystalline materials in fusion reactors caused by helium, which leads to premature failure, often within a year. Here we experimentally demonstrate that nanodispersoids with constitutional vacancy-like atomic-scale free volume can securely store helium, not only at the matrix-dispersoid interface but also within their bulk lattices, which suggests their effectiveness in delaying critical helium damage of the polycrystalline matrix. The selected model nano-heterophase, fayalite Fe2SiO4, possesses moderately strong lattice sinks for helium while undergoing lattice distortions upon helium absorption. These distortions cause observable changes in peak intensities of X-ray diffraction (XRD) patterns, distinct from changes resulting from other factors like radiation damage. By comparing grazing incidence XRD patterns with ab initio computed patterns, we show that such nano-heterophases can store up to ∼10 at% helium within their bulk lattice, forming a “helide compound.” Incorporating just 1 vol% of Fe2SiO4 reduces helium bubble size and number density by >20 % and >50 %, respectively. These findings suggest that 1–2 vol% of appropriate nano-heterophases can accommodate a few thousand appm of bulk helium, expected to be generated over a 10-year operational period.

Evaluation of shielding and interaction properties of different stainless steel alloys for nuclear power plant shielding

This study evaluated the gamma and neutron-shielding parameters, electron and charge particles’ interaction properties of three different stainless steel-based alloys (2507SS, 304SS and AISI1018) with densities of 7.82, 7.81 and 7.82, respectively for deployment in nuclear reactor construction. Gamma shielding parameters were evaluated with Phy-X/PSD, Py-MLBUF and GRASP computer programs within the energy range of 0.2 MeV to 15 MeV. The relative deviations (RD %) between two of these computational platforms were calculated. Fast Neutrons Effective Removal Cross-Sections (FNRCS) and Macroscopic Removal cross-section (MRCS) values for fast neutrons were calculated with Phy-X/PSD and MRCS software, while thermal/fast neutron attenuation factors were calculated with NGCals software. The continuous-slowing-down approximation (CSDA) range and total stopping power (TSP) values of H+ and He++ ions were estimated with SRIM Monte Carlo code in a wide energy range of 1–20 MeV. The projectile range of electrons was computed with ESTAR NIST software within the kinetic energy of 0.01–1000 MeV. Gamma-ray, thermal neutron and fast neutron transmission factors (TFs) values were calculated for all samples for a range of well-known energies. From the obtained results, all evaluated parameters were dependent on the composition of shielding materials, the type of radiation and the photon energy of the radiation. The maximum MAC values obtained from the three alloy samples are 0.1505 cm2/g for 2507SS, 0.1453 cm2/g for 304SS and 0.1459 cm2/g for AISI1018 at 0.02 MeV. These results projected 2507SS as a better shielding alloy when compared with other investigated alloy samples. Considering the closeness of densities of investigated alloy samples the shielding and interaction properties of all alloy samples gave excellent and similar results implying that all investigated alloys will effectively serve as a radiation-shielding material in a nuclear reactor.

Numerical simulation of inductively coupled Ar/O<sub>2</sub> plasma

In the inductively coupled plasma (ICP) discharge, surface processes, such as reflection, de-excitation, and recombination, can occur when active species arrive at material surfaces, which accordingly influences the plasma properties. In this work, a fluid model is used to study the Ar/O<sub>2</sub> plasma generated by ICP reactors made of different materials. In simulation, sticking coefficient is employed to estimate the surface reactions on different materials. As the reactor material changes from stainless steel to anodized aluminum to Cu, the sticking coefficient of surface reaction O→1/2O<sub>2</sub> decreases accordingly. It is found that the reactor material has a great effect on species density. In the stainless steel reactor, the density of O atoms at grounded state and excited state are much lower because more O<sub>2</sub> molecules are generated from the surface reaction, yielding a much higher density of <inline-formula><tex-math id="M5">\begin{document}$ {\text{O}}_2^ + $\end{document}</tex-math><alternatives><graphic specific-use="online" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M5.jpg"/><graphic specific-use="print" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M5.png"/></alternatives></inline-formula> molecular ions which are mainly created from the ionization process of O<sub>2</sub> molecules. Similarly, the high density of O<sub>2</sub> molecules also enhances the production of <inline-formula><tex-math id="M6">\begin{document}${{{\mathrm{O}}} _2}\left( {{{\mathrm{a}}^1}{\Delta _{\mathrm{g}}}} \right)$\end{document}</tex-math><alternatives><graphic specific-use="online" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M6.jpg"/><graphic specific-use="print" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M6.png"/></alternatives></inline-formula> molecules through the excitation process and O<sup>–</sup> ions through the dissociation attachment reaction. On the contrary, more electrons are consumed via the collisions between electrons and O<sub>2</sub> molecules or <inline-formula><tex-math id="M7">\begin{document}$ {\text{O}}_2^ + $\end{document}</tex-math><alternatives><graphic specific-use="online" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M7.jpg"/><graphic specific-use="print" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M7.png"/></alternatives></inline-formula> molecular ions. Therefore, the electron density obtained in the Cu reactor is highest. The density of Ar<sup>+</sup> ions and Ar<sub>m</sub> atoms also increase with sticking coefficient decreasing. The density of O<sup>+</sup> ions and <inline-formula><tex-math id="M8">\begin{document}$ {\text{O}}_2^ + $\end{document}</tex-math><alternatives><graphic specific-use="online" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M8.jpg"/><graphic specific-use="print" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M8.png"/></alternatives></inline-formula> molecular ions peak below the coil in the stainless steel reactor, whereas the radial uniformities are improved in the Cu reactor. In the three reactors, the electrons distribute evenly at the reactor center region. The O density and <inline-formula><tex-math id="M9">\begin{document}${{{\mathrm{O}}} _2}\left( {{{\mathrm{a}}^1}{\Delta _{\mathrm{g}}}} \right)$\end{document}</tex-math><alternatives><graphic specific-use="online" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M9.jpg"/><graphic specific-use="print" xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="13-20240436_M9.png"/></alternatives></inline-formula> density significantly peak at the reactor center, while the maximum value of Ar<sup>+</sup> density and Ar<sub>m</sub> density are below the coil. As for O(<sup>1</sup>D), the maximum density below the coil region moves toward the reactor center as the reactor material changes from stainless steel to Cu. Finally, the effect of sticking coefficient of O→1/2O<sub>2</sub> is studied. The results show that the O atom density decreases with the sticking coefficient increasing, but the opposite trend is observed in O<sub>2</sub> molecular density. It is noticed that the sticking coefficient has little effect on species density when it is higher than 0.5.