Some chemical and microstructural factors influencing creep cavitation resistance of austenitic stainless steels

Creep cavitation in materials is greatly influenced by trace elements. To enhance creep cavitation resistance, the chemical composition of 304, 321, 347 austenitic stainless steels was modified with the addition of minute amounts of boron and cerium. The addition of boron and cerium to type 304 stainless steel led to an increase in its creep rupture life with an associated decrease in creep rupture ductility. The addition of boron and cerium to the titanium-containing 321 steel and niobium-containing 347 steel was found to increase their creep rupture life and ductility. Creep cavitation was highly suppressed in the 347 and 321 steels with the addition of boron and cerium. The chemistry of the grain boundary and creep cavity surface was analyzed by Auger electron spectroscopy. Extensive sulphur segregation was observed on the grain boundary and cavity surface of the steels without boron and cerium addition and even in the 304 steel containing boron and cerium. In the boron- and cerium-containing 347 and 321 steels, respectively, segregation of elemental boron and the BN compound on the cavity surface were observed. These segregations reduced cavity growth rate substantially in these steels and BN segregation was found to be more effective in reducing cavity growth rate than boron segregation. Cerium acts as a getter for soluble sulphur in the steels by precipitation of ceriumoxysulfide (Ce₂O₂S) to facilitate the segregation of boron on the cavity surface.     

超临界水冷堆燃料包壳材料的辐照损伤研究进展

超临界水冷堆（SCWR）是第四代核电站的主力堆型之一,高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类,奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究,包括:镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现,辐照产生的缺陷主要是间隙型位错环,伯格斯矢量为1/3<111>,在较高剂量（5～7 dpa）辐照下,出现空洞肿胀。在氢滞留的影响下,位错环有着独特的演化规律,总结提出了位错环的四阶段演化过程。The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.     

Solute segregation to grain boundaries and free surfaces in an Fe---Si multicomponent alloy

Solute segregation was measured at both the {310} symmetrical tilt grain boundary and the (310) free surface of a sample of an Fe-6at%Si alloy containing traces of P, S, N and C at 873 K. Large phosphorus enrichment and silicon depletion characterize the grain boundary segregation in spite of a different bulk concentration of nitrogen. The surface segregation in nitrogen-containing samples is controlled by strong cosegregation of Si and N, resulting in the formation of a stable Si_xN_y 2D surface compound, whereas pronounced surface segregation of sulphur dominates in denitridized samples. The differences of grain boundary and surface segregation are discussed as a kind of “anisotropy of interfacial segregation” on the basis of Guttmann's theory with different values of free energies of segregation to grain boundary and free surface. They also suggest that the measurements of surface segregation cannot be unambiguously used for predicting the grain boundary segregation. In some non-brittle multicomponent systems, a better way of predicting segregation behavior at grain boundaries would be the measurement of grain boundary segregation in a related system with solute concentrations that cause embrittlement. The findings can then be applied to the required alloy composition on the basis of Guttmann's theory.     

Non-equilibrium grain boundary cosegregation of Mo and P

The segregation of P at grain boundaries is believed to be an important cause of temper embrittlement in steels. As an alloy element, Mo may reduce the embrittlement. However, the concentration measured by Auger electron spectroscopy at the grain boundary in 2.25Cr1MoV and 12Cr1MoV showed that the concentration of P increased with that of Mo, which indicates that Mo and P cosegregated to the grain boundary in Cr-Mo steels.     

Effects of excess point defects flow and interface orientation on grain boundary migration under irradiation

Fe-Cr-Ni, Ni-Al and Ni-Si alloys, and 316L stainless steels as reference were electron-irradiated using a high voltage electron microscope (1MV), and in-situ observations of structural evolution and micro-chemical analysis were carried out. From the compositional analysis it was found that nickel was enriched and chromium depleted near grain boundary in Fe-Cr-Ni alloys including 316L stainless steels, and that simultaneously grain boundary migration was caused during irradiation, even if no grain boundary migration occurred in the un-irradiated area at the same irradiation temperature. The occurrence of boundary migration strongly depended upon orientation relationship between boundary interfaces. It is suggested that grain boundary migration under irradiation remarkably occurs in the alloys in which solute enrichment is taken place at the grain boundary as a result of the flow of radiation-introduced point defects into grain boundary and that their magnitude depend upon net flow of point defect, especially that of under-sized interstitial atoms.     

Micromechanistic origin of irradiation-assisted stress corrosion cracking

A multi-scale study of the micromechanics of dislocation–grain boundary interactions in proton and ion-irradiated stainless steels is presented. Interactions of dislocation channels with grain boundaries result in slip transfer, discontinuous slip without or with slip along the grain boundary. The presence of the irradiation damage enhances the importance of the magnitude of the resolved shear stress on the slip system activated by the grain boundary to transfer slip across it. However, the selected slip system is still determined by the minimization of the grain boundary strain energy density condition. These findings have implications for modelling the mechanical properties of irradiated metals as well as in establishing the mechanism for disrupting the grain boundary oxide, which is a necessary prerequisite for irradiation-assisted stress corrosion cracking.     

Effects of B segregation on Mo-rich phase precipitation in S31254 super-austenitic stainless steels: Experimental and first-principles study

Precipitation in super-austenitic stainless steels will significantly affect their corrosion resistance and hot workability. The effects of Cr and Mo on precipitation behaviors were mainly achieved by affecting the driving force for precipitation, especially Mo has a more substantial promotion effect on the formation of the σ phase than Cr. In the present study, B addition to the S31254 super-austenitic stainless steels shows an excellent ability to inhibit precipitation. The effect of B on the precipitation behaviors was investigated by microstructure characterization and theoretical calculations. The experimental observation shows that the small addition of B inhibits the formation of the σ phase along grain boundaries and changes from continuous to intermittent distribution. Moreover, the inhibitory effect increased obviously with the increase of B content. The influence of B addition was theoretically analyzed from the atomic level, and the calculation results demonstrate that B can inhibit the formation of σ phase precipitates by suppressing Mo migration to grain boundaries. It is found that B and Mo are inclined to segregate at Σ 5 and Σ 9 grain boundaries, with B showing the most severe grain boundary segregation tendency. While B distribution at the grain boundary before precipitation begins, the segregation of Mo and Cr will be restrained. Additionally, B's occupation will induce a high potential barrier, making it difficult for Mo to diffuse towards grain boundaries.     

高He通量注入条件下F/M和ODS钢中纳米微结构对He泡形核和长大的影响

利用LEAF装置提供的2 MeV的He离子,在500和600 °C分别对新型F/M钢-SIMP钢和ODS钢(MA956和Eurofer-ODS钢)注入1×1017 ions/cm2的高通量He离子,借助透射电子显微镜,表征了辐照后三种材料的肿胀行为,验证了各材料中纳米微结构(晶界,析出相和纳米氧化物)对辐照后He泡成核和长大的影响。结果表明,基于材料中晶界和析出相对He泡生长的抑制作用,温度为500 °C时,SIMP和Eurofer-ODS钢表现出较高的抗辐照肿胀性能,而MA956中纳米界面He泡成核和长大作用不明显,表现出较差的抗辐照肿胀性能;此外,温度为600 °C时,Eurofer-ODS钢由于其晶界和氧化物界面的较强作用,表现出较好的抗辐照肿胀性能。总体来说,在高He通量注入条件下,材料中纳米结构的存在会抑制He泡长大的过程,但不同材料中纳米结构对He影响作用不同。     

The atkermal stress component in neutron irradiated mild steel

Abstract

The temperature dependence of the tensile lower yield stress of an annealed aluminium grain size controlled mild steel has been investigated in the range 23–250 °C and at a strain rate of 1.67 × 10^?4 sec^?l before and after neutron irradiation to 2.3 × 10¹⁸ n/cm² (fission). The yield stress of the irradiated steel decreases with increasing temperature due to thermal activation of the radiation damage and is predicted to reach asymptotically that of the unirradiated steel at ～285 °C; the maximum test temperature was below that at which thermal annealing of the damage occurs. This implies that the athermal stress component due to irradiation is zero and hence there is negligible long range interaction between dislocations and radiation-induced defects.     

Diffusion of nickel in single- and polycrystalline Cr2O3

Chromia protective layers are formed on many industrial alloys to prevent corrosion by oxidation. Their role is to limit the inward diffusion of oxygen and the outward diffusion of cations. A number of chromia-forming alloys contain nickel as a component, such as steels, FeNiCr and NiCr alloys. To ascertain if chromia is a barrier to outward diffusion, nickel diffusion in chromia was studied in both single crystals and polycrystals in the temperature range 900–1100°C at an oxygen pressure of 10^?4 atm (argon + 100 ppm O₂). A nickel film of ～35 nm thick was deposited on the chromia surface and, after diffusing treatment, nickel penetration profiles were established by secondary ion mass spectrometry (SIMS). Two diffusion domains appear in polycrystals, the first domain is assigned to bulk diffusion and the second is due to diffusion along grain boundaries. For the bulk diffusion domain and diffusion in single crystals, using a solution of Fick's second law for diffusion from a thick film, bulk diffusion coefficients were determined at 900 and 1000°C. At the higher temperature, a solution of Fick's second law for diffusion from a thin film could be used. For the second domain in polycrystals, Le Claire's model allowed the grain boundary diffusion parameter (αD _gb δ) to be established. Nickel bulk diffusion does not vary significantly according to the microstructure of chromia. The activation energy of grain boundary diffusion is slightly greater than the activation energy of bulk diffusion, probably on account of segregation phenomena. Nickel diffusion was compared with cationic self-diffusion and with literature data on Fe and Mn heterodiffusion in the bulk and along grain boundaries. All results were analyzed in relation to the oxidation process of stainless steel.     

Effect of irradiation on microstructure and hardening of Cr–Fe–Ni–Mn high-entropy alloy and its strengthened version

ABSTRACT

A single-phase fcc high-entropy alloy (HEA) of 20%Cr–40%Fe–20%Mn–20%Ni composition and its strength with yttrium and zirconium oxides version was irradiated with 1.4?MeV Ar ions at room temperature and mid-range doses from 0.1 to 10 displacements per atom (dpa). Transmission electron microscopy (TEM), scanning transmission electron microscopy with energy dispersive X-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterise the radiation defects and microstructural changes. Nanoindentation was used to measure the ion irradiation effect on hardening. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages, a comparison was performed with hardening behaviour of 316 austenitic stainless steel irradiated under an identical condition. It was shown that hardness increases with irradiation dose for all the materials studied, but this increase is lower in high-entropy alloys than in stainless steel.     

Grain boundary impurity segregation and neutron irradiation effects in ferritic alloys

Segregation of alloying and impurity elements to grain boundaries in ferritic steels and alloys is known to modify the mechanical properties. This paper considers segregation of such elements, in particular phosphorus and carbon, that occur in ferritic nuclear pressure vessel steels subject to neutron irradiation and temperature typical of that encountered in service. Models are presented that allow the prediction of equilibrium and non-equilibrium segregation of phosphorus to grain boundaries and also take into account synergistic interaction with carbon under various combinations of neutron-irradiation temperature. These are related to a wide range of experimental observations compiled from data in the literature for mainly phosphorus and carbon measured at grain boundaries in neutron-irradiated ferritic vessel steels and alloys. The predictions from the segregation models are compared with these experimental data. The discussion provides a rationalization for the apparent variability in the measured grain boundary phosphorus compositions and thereby fracture susceptibility for various nuclear pressure vessel ferritic steels.     

Influence of nitrogen on the mechanism of austenitic stainless steel hardening

On the example of a C18N12M2 austenitic stainless steel, the influence of nitrogen (whose content varied from 0 to 0.45 wt.%) on the grain boundary hardening coefficient k _h entering into the Hall-Patch equation is analyzed. High values of k _h in steels with and without nitrogen are found. The data of the Auger analysis show that the hardening coefficient in the steel without nitrogen is determined by the grain-boundary segregation of carbon and oxygen. The grain-boundary hardening in the steel with nitrogen is not connected with the predominant segregation of nitrogen at grain boundaries. It is completely governed by intragranular processes—interaction of nitrogen atoms with dislocations. Omsk State Pedagogical University. Translated from Izvestiya Vysshikh Uchebnykh Zavedenii, Fizika, No. 7, pp. 47–52, July, 1999.     

聚变堆候选金属材料的惰性气体离子辐照损伤的研究

综述了有关核聚变反应堆材料的辐照损伤问题的研究,主要包括国产316L奥氏体不锈钢中氦的扩散与氦泡形核生长的研究、316L及低活化FeCrMn合金的高能Ar离子辐照缺陷与空洞肿胀的研究、近期开展的低活化马氏体钢和氧化物颗粒弥散强化合金的高能Ne离子辐照损伤和效应的研究成果。This paper gives a review of our recent studies on the irradiation damage induced by energetic inert-gasions in metallic materials candidate to fusion reactors. The work includes the study of helium diffusion and helium bubble formation in 316L stainless steels, the study of void formation and swelling in the low-activation Fe-Cr-Mn alloy irradiated with high-energy Ar ions, the study of irradiation damage in some low-activation Fe-based steels and ODS alloys by high-energy Ne ions.     

Microstructure evolution of T91 steel after heavy ion irradiation at 550 ℃

Fe-Cr ferritic/martensitic (F/M) steels have been proposed as one of the candidate materials for the Generation IV nuclear technologies. In this study, a widely-used ferritic/martensitic steel, T91 steel, was irradiated by 196-MeV Kr⁺ ions at 550 ℃. To reveal the irradiation mechanism, the microstructure evolution of irradiated T91 steel was studied in details by transmission electron microscope (TEM). With increasing dose, the defects gradually changed from black dots to dislocation loops, and further to form dislocation walls near grain boundaries due to the production of a large number of dislocations. When many dislocation loops of primary a₀/2<111> type with high migration interacted with other defects or carbon atoms, it led to the production of dislocation segments and other dislocation loops of a₀<100> type. Lots of defects accumulated near grain boundaries in the irradiated area, especially in the high-dose area. The grain boundaries of martensite laths acted as important sinks of irradiation defects in T91. Elevated temperature facilitated the migration of defects, leading to the accumulation of defects near the grain boundaries of martensite laths.     

Positron annihilation studies of heavy-ion induced radiation damages in stainless steels

Variable-energy (0–30 keV) positrons were used to study the depth distribution of heavy-ion induced vacancy type defects in the following specimens, SUS304 and SUS316 austenitic stainless steels, a SUS444 ferritic stainless steel and nickel metal, which were irradiated by 0.5 MeV He⁺, 2.0 MeV C⁺, 3.5 MeV Si²⁺ and 4.0 MeV Ni²⁺ ions up to 0.01 dpa at peak. Vacancy type effective residual defects (ERD) were evaluated from the line shape parameterS of Doppler broadened positron annihilation photon spectra. With increasing primary knockon atom (PKA) energy, a decrease of the vacancy type ERD was observed. The ERD differences among the specimens are discussed in comparison with theoretical predictions.     

高合金钢光谱分析准确度改进试验

在我国钢铁厂中,高合金钢的光谱分析受到准确度的限制,没有普遍展开。通过对两种有代表性的高合金钢,即P18高速钢及Я1不锈钢的试验,证明用摄谱方法采取种种改进以后,高合金钢的分析约可达到最小的百分均方误差1％左右。文中对某些工厂的先进经验曾加以介绍。对于高合金钢的准确度问题曾予讨论。     

Helium effects on the microstructural evolution of reactor irradiated ferritic and austenitic steels

Abstract

Microstructural evolution in neutron irradiated austenitic stainless steels and Cr-Mo ferritic steels is reviewed. Important highlights are: (1) there is a strong correlation between precipitation and void evolution in austenitic steels; (2) helium affects precipitate evolution in austenitic steels, but observations indicate no effect on precipitation in ferritic steels; (3) helium has a pronounced effect on the cavity evolution of the two steel types. Helium effects are explained in terms of the interrelationship between microstructural evolution and point-defect annihilation processes. In stainless steel, three relative regimes of microstructural behavior for different helium generation rate-displacement rate ratios are recognized: (1) “low” He/dpa ratio, where helium effects on the radiation-induced microstructural evolution are negligible or develop slowly, (2) “medium” He/dpa ratio, where helium effects strongly enhance the microstructural changes, and (3) “high” He/dpa ratio, where helium effects are limited to the early development of a high density of fine bubbles which interfere with other radiation-induced microstructural changes, but allow enhanced thermal microstructural evolution to take place instead. The extensive data on austenitic steels fall within these regimes. Ferritic steels are known to be highly resistant to void swelling without helium. It is suggested that enhanced cavity formation due to helium in ferritic steels makes higher swelling a potential concern for fusion reactor applications.     

Evolution of steel surface composition with heating in vacuum and in air

X-ray photoelectron spectroscopy (XPS) has been used to investigate the changes in surface composition of three steels as they have undergone heating. The steels were mild steel, and two austenitic stainless steels, commonly designated 304 and 316 stainless steels. XPS measurements were made on the untreated samples, and then following heating for 30 min in vacuo and in a 1 × 10⁻⁶ Torr partial pressure of air, at temperatures between 100 °C and 600 °C.Mild steel behaves differently to the two stainless steels under the heating conditions. In mild steel the iron content of the surface increased, with oxygen and carbon decreasing, as a function of increasing temperature. The chemical state of the iron also changed from oxide at low temperatures, to metallic at temperatures above 450 °C.In both stainless steels the amount of iron present in the surface decreased with increasing temperature. The decrease in iron at the surface was accompanied by an increase in the amount of chromium at the steel surface. At temperatures above 450 °C the iron in both 304 and 316 stainless steels showed significant contributions from metallic iron, whilst the chromium present was in an oxide state. In 316 stainless steel heated to 600 °C there was some metallic chromium present in the surface layer.The surfaces heated in air showed the least variation in composition, with the major change being the loss of carbon from the surfaces following heating above 300 °C. There was also a minor increase in the concentration of chromium present on both the stainless steels heated under these conditions. There was also little change in the oxidation state of the iron and chromium present on the surface of these steels. There was some evidence of the thickening of the surface oxides as seen by the loss of the lower binding energy signal in the iron or chromium core level scans.The surfaces heated in vacuum showed a similar trend in the concentration of carbon on the surfaces, however the overall concentration of oxygen decreased throughout the heating of these steels. There were also significant changes in the oxidation state of the iron and chromium on these surfaces with significant amounts or iron and chromium present in the metallic form following heating up to 600 °C.It appears that the carbon contamination on the surfaces plays an important role in the fate of the surface oxide layer for all of the steels heated in a vacuum environment.