9Cr-ODS Steel Research Articles

Early corrosion behavior of low carbon 9Cr-ODS steel in high temperature oxygen-saturated lead-bismuth eutectic

The early corrosion behavior of low carbon 9Cr-ODS steel in 500 °C oxygen-saturated LBE was investigated. The thickness of the oxide film was found to increase with the duration of corrosion. Results from Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM), and the Shanghai Synchrotron Radiation Facility (SSRF) indicate that the oxide layer on 9Cr-ODS steel is multilayered, consisting of a (Fe, Cr)3O4 spinel inner layer, a nanocrystalline transition layer rich in Cr and Mn but poor in Fe, and an Fe3O4 outer layer. The corrosion mechanism of low carbon 9Cr-ODS steel in liquid LBE is discussed.

Corrosion Behavior of Ferritic 12Cr ODS and Martensitic X46Cr13 Steels in Nitric Acid and Sodium Chloride Solutions.

This paper presents corrosion resistance results of a 12Cr ferritic ODS steel (Fe-12Cr-2W-0.5Zr-0.3Y2O3) fabricated via a powder metallurgy route as a prospective applicant for fuel cladding materials. In a spent nuclear fuel reprocessing facility, nitric acid serves as the primary solvent in the PUREX method. Therefore, fundamental immersion and electrochemical tests were conducted in various nitric acid solutions to evaluate corrosion degradation behavior. Additionally, polarization tests were also performed in 0.61 M of sodium chloride solutions (seawater-like atmosphere) as a more general, all-purpose procedure that produces valid comparisons for most metal alloys. For comparison, martensitic X46Cr13 steel was also examined under the same conditions. In general, the corrosion resistance of the 12Cr ODS steel was better than its martensitic counterpart despite a lower nominal chromium content. Potentiodynamic polarization plots exhibited a lower corrosion current and higher breakdown potentials in chloride solution in the case of the ODS steel. It was found that the corrosion rate during immersion tests was exceptionally high in diluted (0.1-3 M) boiling nitric acid media, followed by its sharp decrease in more concentrated solutions (>4 M). The results of the polarization plots also exhibited a shift toward more noble corrosion potential as the concentrations increased from 1 M to 4 M of HNO3. The results on corrosion resistance were supported by LSCM and SEM observations of surface topology and corrosion products.

Oxide particles in oxide dispersion strengthened steel neutron-irradiated up to 158 dpa at Joyo

We investigated the stability of oxide nano particles in oxide dispersion-strengthened (ODS) steel, which is a promising candidate material for next-generation reactors, under neutron irradiation at high temperature to high doses. MA957, a 14Cr-ODS steel, was irradiated with Joyo in Japan Atomic Energy Agency under irradiation conditions of 130 dpa at 502 ºC, 154 dpa at 589 ºC, and 158 dpa at 709 ºC. Three-dimensional atom probe (3D-AP) and transmission electron microscope (TEM) observation were performed to characterize the oxide particles in the ODS steels. A high number density of Y-Ti-O particle was observed in the unirradiated and irradiated samples. Almost no change in the morphology of the oxide particles, i.e. average diameter, number density, and chemical composition, has been observed in the samples irradiated to 130 dpa at 502 ºC and to 154 dpa at 589 ºC. A slight decrease in number density was observed in the sample irradiated to 158 dpa at 709 ºC. The hardness of any of the irradiated samples was almost unchanged from that of the unirradiated sample. It was revealed that the oxide particles existed stable, and the strength of the material was sufficiently maintained even after being neutron irradiated to high dose of ∼160 dpa at high temperature up to 700 ºC.

Effect of annealing temperature on the evolution of microstructure, texture, and mechanical properties of hot-rolled 12Cr-ODS steel

To achieve high thermal efficiency, today's nuclear reactor structures are exposed to higher temperatures and pressures, which requires the use of high-strength steels with specific properties, such as ODS steels. There is a need to clarify the evolution of the microstructure and properties of steels at elevated temperatures. This study systematically investigates the evolution of microstructure, texture, and mechanical property variations of 12Cr-ODS steels after hot-rolling and subsequent annealing at 1000 °C and 1200 °C. The investigation utilizes optical microscopy, electron backscatter diffraction technique, and mechanical property measurements. The microstructure of hot-rolled samples shows a layered alternating distribution, which is distinguished by a notable presence of the α-fiber texture. Following annealing at 1000 °C, the martensite grains are smaller with a reduced hardness, while maintaining a strong α-fiber texture. After annealing at 1200 °C, there is a rapid increase in the growth of martensite grains, a significant rise in hardness, a reduction in the α-fiber texture characteristics, and an improvement in the γ-fiber texture characteristics. Moreover, the maximum intensity of the α-fiber texture diminishes as the annealing temperature increases. The mechanical properties of the samples deteriorated after annealing at 1200 °C, which can be attributed to the coarse martensite grains and the texture components containing the {001} cleavage plane dominating the occurrence of brittle cleavage fracture. The 0.1Y sample after annealing at 1000 °C exhibits an excellent combination of strength (1458 MPa) and ductility (20.3%), which is owing to the unique heterogeneous grain structure and the evolution of favorable texture.

Corrosion resistance of 9–12Cr ODS steels in pressurized water for fusion application

The blanket design of Chinese fusion engineering test reactor (CFETR) may utilize pressurized water as coolant. The fusion structural materials will endure high temperature, high pressure and result in corrosion in the pressurized water. It is necessary to verify corrosion resistance of the materials. In the present study, three kinds of advanced fusion structural materials with excellent mechanical properties and neutron irradiation resistance, i.e., 9Cr-ODS steel, 9CrODS-Al steel, and 12CrODS-Al steel were immersed in pressurized water at 325 °C under the pressure of 15.5 MPa up to 859 h. Room-temperature tensile properties were investigated for the materials before and after corrosion. The oxidation layers of the steels were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The ODS steels have good corrosion resistance in the pressurized water, especially the 9CrODS-Al and 12CrODS-Al steels. The outer oxide layer is identified as Fe2O3 for all the steels. The inner layer is mainly FeCr2O4 for the 9CrODS steel, and Fe(Al,Cr)2O4 for the ODS steels with Al element which better protect the steels from further corrosion. The relevant properties data can provide reference for the blanket materials choice of CFETR and the future fusion reactors.

Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C

ABSTRACT The Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650°C–850°C. However, little data have been obtained above 850°C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700°C–1000°C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix’s phase transformation, and a single equation can express a creep rupture strength from 700°C to 1000°C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.

Dependency of Portevin-LeChatelier Effect and Microstructure Characteristics of 9Cr ODS Steels on Strain Rates Upon Hot Deformation

Dependency of Portevin-LeChatelier Effect and Microstructure Characteristics of 9Cr ODS Steels on Strain Rates Upon Hot Deformation

Comparison of Corrosion Behavior of T91, 9Cr and 9CrAl ODS Steels in Liquid Pb

It is important to improve the liquid lead corrosion resistance of fuel cladding alloy for promoting the development of lead-based reactors. The corrosion behaviors of traditional T91 steel and similar oxide dispersion strengthen ODS-type steels with or without the addition of Al, and were examined and compared at 600 °C in static oxygen-controlled liquid Pb in this research. High-temperature liquid lead corrosion tests were carried out for 120 h, 240 h, 500 h, 1000 h, and 2000 h, respectively, for three prepared samples. After the experiment, the corrosion behavior was evaluated and compared mainly based on the aspects of appearance, corrosion depth, microstructure, and composition difference. It was found that just the ODS design did not show a positive effect on corrosion resistance, while the addition of Al is beneficial to improving the corrosion resistance of ODS steel. The maximum corrosion depth of 9CrAl ODS is only 51.8 μm after corrosion for 2000 h, which is much lower than that of 9Cr-ODS steel. A thin film containing Al/Cr formed in the corrosion area after adding Al in 9Cr ODS steel, which played a positive role in corrosion resistance.

Microstructure morphology and tensile properties of 9Cr-ODS steels added different Y2O3 contents

Oxide dispersion strengthened (ODS) steels are considered to be the promising structural material candidates for Generation IV fission reactors, fusion reactor applications, new accident-tolerant fuel (ATF) cladding materials. In this study, different contents of Y2O3 were added into 9Cr-ODS steels to realize different Y/Ti atomic ratios. The microstructure of samples, kinds, number densities and size distribution characteristic of nano-scale oxide-precipitates formed in the 9Cr-ODS steel samples were investigated by synchrotron radiation SAXS, XAFS spectrum and TEM, explained the effect of Y2O3 contents on microstructures of 9Cr-ODS steel samples. The tensile properties of the 9Cr-ODS steel samples with different Y2O3 contents were also measured simultaneously. The experimental data show when Y2O3 content is 0.3%, the density of Y-Ti-O nano-clusters is the highest, reaching to 1.39×1024 /m3. At the same time, the number density of Y-Ti-O nano-clusters is the lowest, if the Y2O3 content is 0.7%. The XAFS result of 0.7% Y2O3 sample is close to a Y2O3 reference sample. The 0.3% Y2O3 has a bigger grain size than the other two samples. The 9Cr-ODS steels with Y2O3 content of 0.3% exhibit the highest tensile strength and higher ductility properties.

The effect of Si increase on microstructure and mechanical properties of multi-nano-oxide dispersion strengthened steel

Two 12Cr ODS steels with nominal compositions of Fe-12Cr-1Al-0.5Zr-0.35Y2O3-xSi (x = 1 and 2 wt%, respectively) were successfully fabricated. In both steels, a high number density of nanoparticles was produced as dispersion strengthening precipitates. These nanoparticles were identified to be nano-oxides of various types of Y-Al-O, Y-Si-O, and Y-Zr-O, and each type exhibited a different feature size. Increasing the Si content from 1 to 2 wt% brought a twofold benefit, by further refining matrix grains and producing more ultra-fine Y-Si-O and Y-Zr-O nanoparticles. In addition, the mechanical strengths were improved without the sacrifice of plasticity at 973 K.

Effects of grains‘ morphology on strengthening mechanisms in ODM401 14Cr ODS steel at high temperatures

The effect of grain orientation on the mechanical behavior of the 14Cr ODS steel (ODM401) at temperatures 400–800 °C is analyzed. The tensile and stress relaxation tests are performed on tensile specimens with grain orientation parallel and perpendicular to the extrusion direction. The strengthening mechanisms are analyzed in terms of the internal and thermal stress component, dislocation glide and climb and grain boundary creep. Models of strengthening mechanisms are used for evaluation of their intensity for a given temperature, grain orientation and macroscopic plastic strain level (0.001, 0.01, 0.025). Both grain orientations provide similar strengthening up to 600 °C. The grain orientation parallel to the extrusion direction provides higher strength at temperatures 700–800 °C due to the reduced grain boundary creep and dislocations annihilation at grain boundaries.

A model experiment on nano-particle stability in 12Cr-ODS steel using high-resolution high voltage microscopy

In this work, the microstructure stability of 12Cr-ODS ferritic steel under irradiation is studied using a high-resolution high voltage electron microscope up to 1.5 dpa at 300 and 500 °C. Coherent YTiO3 nanoparticles (NPs) undergo coarsening during electron irradiation and the coarsening rate is positively correlated to irradiation temperature (the third power of NPs size is proportional to the irradiation dose). He pre-implantation can significantly reduce the coarsening rate. A model coupling Ostwald ripening and irradiation enhanced diffusion is used to explain this coarsening. The fitting results between NPs size and irradiation dose show that pre-implanted He causes a lower diffusion of Y in 12Cr-ODS steel compared with the non-implanted sample. In response to the irradiation, plate-like radiation-induced precipitates appear which result in diffraction streaks in fast fourier transform (FFT) pattern. Precipitations are paralleled to [100] or [110] direction of Fe matrix and located near the interface between NPs and matrix. The hypothesis that coherent Cr precipitate along NPs/matrix interface caused diffraction streaks is proposed. Besides, He pre-implantation can also delay this precipitation.

Effect of post-weld heat treatment on a friction stir welded joint between 9Cr-ODS and CLF-1 steels

Dissimilar-metals welding is essential to construction of a fusion reactor by using advanced blanket structural materials of oxide dispersion strengthened (ODS) steels. In the present study, Dissimilar-metals bonding between 9Cr-ODS and CLF-1 steels was carried out by friction stir welding (FSW). Effect of post-weld heat treatment (PWHT) on microstructure and bonding properties of the joint was investigated. The results show that, quenched martensite was found in the weld metal (WM) after FSW. PWHT at 740-800 °C can recover the microstructure of the WM to tempered martensite. Mid-size particles and nano-particles were observed and analyzed in the WM. Besides, in the WM on the CLF-1 side, recrystallization was found after PWHT. As the PWHT temperature increased, the recrystallization width and grain size grew up. However, the recrystallization in the WM did not degrade but may be beneficial to the bonding properties of the joint due to support ductile deformation during mechanical tests. The recommend PWHT temperature is 740-760 °C for keeping the mechanical properties of the CLF-1 base metal (BM) as that before FSW.

The microstructure evolution and irradiation hardening in 15Cr-ODS steel irradiated by helium ions

Oxide dispersion strengthened (ODS) steels with excellent irradiation swelling resistance and high temperature creep resistance are regarded as one of the priority candidate materials for the first wall/blanket materials in fusion reactors. Microstructure evolution and irradiation hardening in 15Cr-ODS steel irradiated by helium ions at R.T. and 400 °C respectively have been studied by transmission electron microscopy and nanoindentation technique. Three main oxides were identified, including high-density dispersed Y-Ti-O nanoclusters, tens of nanometers Y2TiO5 and a few hundred nanometers Y2Ti2O7 particles with the largest size but the least density. Helium bubbles were homogeneously distributed in the matrix in both irradiated specimens and the aggregation state of bubbles near GBs was related to the nature of GB. The average size of bubbles increased and the density decreased in the steel irradiated at elevated temperature. A defect-depleted zone was observed near a GB with a Y2TiO5 precipitate imbedded and a large helium bubble was present at the interface of the Y2TiO5 particle/GB. It suggests that the GBs combined with Y2TiO5 can enhance the aggregation of helium bubbles since GBs can act as defect sinks as well as diffusion channels and Y2TiO5 has a high ability of trapping helium atoms. Significant irradiation hardening was obtained at both irradiation temperatures. The relationship between microstructure and the irradiation hardening was discussed.

Effects of Grains‘ Morphology on Strengthening Mechanisms in 14cr Ods Steel at High Temperatures

Effects of Grains‘ Morphology on Strengthening Mechanisms in 14cr Ods Steel at High Temperatures

Small-angle neutron scattering study on the stability of oxide nanoparticles in long-term thermally aged 9Cr-oxide dispersion strengthened steel

A 9Cr-oxide dispersion strengthened (ODS) steel was thermally aged at 873 K for up to 5000 h. The size distribution and chemical composition of the dispersed oxide nanoparticles were analyzed by small-angle neutron scattering under a magnetic field. Combined with transmission electron microscopy, Vickers micro-hardness tests and electron backscattered diffraction measurements, all the results showed that the thermal treatment had little or no effect on the size distributions and volume fractions of the oxide nanoparticles in the ferromagnetic matrix, which suggested excellent thermal stability of the 9Cr-ODS steel.

Response of nanoclusters to heavy-ion irradiation in an Fe-12Cr ODS steel

Oxide dispersion-strengthened (ODS) steel is one of the most prominent candidates as the fuel cladding or blanket component in advanced nuclear reactors. The stability and evolution of high-density nanoscale particles in ODS steels during irradiation are of considerable importance in influencing the mechanical properties and irradiation resistance. The microstructure, especially high-density nanoparticles in 12Cr-ODS steel, was investigated in the present study after 1.6 MeV Fe13+ ions irradiation at 450 °C. Atomic probe tomography (APT) and transmission electron microscopy (TEM) were used to track the evolution of microstructure at an atomic scale. Results demonstrate the uniform dispersion of numerous Y-Ti-O nanoclusters in the matrix. The average diameter and Y:Ti ratio of the clusters increased after irradiation. The interface-favorable precipitation of irradiation-induced nanosized α′ phase was especially demonstrated. Radiation-enhanced diffusion (RED) combined with Oswald ripening were considered to illustrate the evolution of Y-Ti-O nanoclusters during irradiation. The segregation behavior of α′ was discussed on the basis of the RED accelerated kinetics and thermodynamic.

Influence of Al addition on the microstructure and mechanical properties of Zr-containing 9Cr-ODS steel

Abstract Two oxide dispersion strengthened (ODS) steels with compositions of 9Cr-ODS (9Cr-ODS) and 9Cr-3Al-ODS (9Cr-ODS-Al) were fabricated by hot isostatic pressing (HIP). The microstructures and precipitates of the ODS samples were characterized by scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), and transmission electron microscopy (TEM). The results showed that the addition of Al could reduce the precipitation of coarse M23C6 and optimized the size of carbides in ODS steels. Furthermore, it was found that the mean size of nanoparticles in 9Cr-ODS-Al is larger than that in 9Cr-ODS due to the formation of large-size Y–Al–O nanoparticles with Al addition. The tensile test results indicated that the addition of Al slightly decreased the strength of ODS steels, but substantially increased the ductility.

A Preliminary Work on the Preparation for Neutron Irradiation of Advanced Fusion Materials Using Small Samples in China

A variety of new materials have been developed during the past two decades in China to support the design and construction of key components for next step fusion devices. Most recently, a domestic project, promoted by the engineering design of China Fusion Engineering Test Reactor (CFETR), has been launched to preliminarily investigate the performance of advanced fusion materials after neutron irradiation. The goal of this paper is to introduce the preparation for the coming neutron irradiation campaign. The development status and acceptance of some outstanding fusion materials are briefly reviewed with the main description focusing on some high-performance 9Cr-ODS steels, Cu–0.3Al2O3 alloys, W-K and W-ZrC alloys for blanket/divertor components. The principles for screening suitable materials out from several candidates are suggested, which consider low activation, sufficient production scale, homogeneous microstructures, safety margin and repeatability of basic properties as the basic requirements. As a recommended approach for neutron irradiation tests, the small sample testing technology (SSTT) is highlighted in this paper. Further development of SSTT using inverse finite element method (FEM) and energy equivalent principle is performed for new designs of small tensile samples.

Effects of Heat Treatment on Microstructure and Mechanical Properties of a Bimodal-grain Ultro-Low Carbon 9Cr-ODS Steel

Effects of Heat Treatment on Microstructure and Mechanical Properties of a Bimodal-grain Ultro-Low Carbon 9Cr-ODS Steel