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Abstract
Fatigue specimens of A508-3 steel were irradiated in the swimming-pool test reactor in China Institute of Atomic Energy, the fluence was 3×10¹⁹ n/cm² at 300°C, then low-cycle fatigue tests were carried out at ambient temperature, with the fatigue strain range is 0.32-1.8%. The results indicate that, irradiated A508-3 specimens exhibit cyclic softening and instability behavior during the test, and the cyclic softening rate increased with strain range increased; fatigue life decreased from 1.7×10⁵ to about 5×10², as the strain range increased from 0.32% to 1.8%, the fatigue life of A508-3 steel increased after the neutron irradiation; fatigue fracture initiated at the surface of specimen, and more individual cracks formed on the specimens of higher strain range compared with the specimens of lower strain range.
... However, the effect of irradiation is not generally considered in fatigue analyses of reactor structures. Some studies have carried out fatigue tests under different irradiation test conditions but obtained different (even contradictory) test conclusions [16][17][18][19]. Under certain irradiation test conditions, the fatigue life of the sample without irradiation is 1.5-2.5 times that of the sample with irradiation [16,17]. ...
... Under certain irradiation test conditions, the fatigue life of the sample without irradiation is 1.5-2.5 times that of the sample with irradiation [16,17]. However, Zhong [18] found that the fatigue life of an irradiated sample was three times greater than that of a non-irradiated sample. Another study concluded that irradiation was beneficial to fatigue life at low strain amplitudes, but not at high strain amplitudes [19]. ...
The influence of irradiation should be considered in fatigue reliability analyses of reactor structures under irradiation conditions. In this study, the effects of irradiation hardening and irradiation embrittlement on fatigue performance parameters were quantified and a fatigue life prediction model was developed. Based on this model, which takes into account the cumulative effect of a neutron dose, the total fatigue damage was calculated according to Miner's linear cumulative damage law, and the reliability analysis was carried out using the Monte Carlo simulation method. The case results show that the fatigue life acquired by taking into account the cumulative effect of irradiation was reduced by 24.3% compared with that acquired without considering the irradiation effect. Irradiation led to the increase of the fatigue life at low strains and its decrease at high strains, which is in accordance with the findings of an irradiation fatigue test. The rate of increase in the fatigue life decreased gradually with the increase of the neutron dose. The irradiation performance parameters had a small influence on fatigue reliability, while the fatigue strength coefficient and the elastic modulus had a great influence on the fatigue reliability. Compared with the current method, which uses a high safety factor to determine design parameters, a fatigue reliability analysis method taking into account the cumulative effect of irradiation could be more accurate in the reliability analysis and life prediction of reactor structures.
... RPV materials were constantly exposed to high temperature and high pressure during the service. They determined the safety and service life of nuclear power plant to a great extent [27][28][29][30]. The nuclear reactor pressure vessel contained the reactor core to prevent the leakage of radioactive substances. ...
The nuclear reactor pressure vessel is an important component of a nuclear power plant. It has been used in harsh environments such as high temperature, high pressure, neutron irradiation, thermal aging, corrosion and fatigue for a long time, which puts forward higher standards for the performance requirements for nuclear pressure vessel steel. Based on the characteristics of large size and wall thickness of the nuclear pressure vessel, combined with its performance requirements, this work studies the problems of forging technology, mechanical properties, irradiation damage, corrosion failure, thermal aging behavior and fatigue properties, and summarizes the research progress of nuclear pressure vessel materials. The influencing factors of microstructures evolution and mechanism of mechanical properties change of nuclear pressure vessel steel are analyzed in this work. The mechanical properties before and after irradiation are compared, and the influence mechanisms of irradiation hardening and embrittlement are also summarized. Although the stainless steel will be surfacing on the inner wall of nuclear pressure vessel to prevent corrosion, long-term operation may cause aging or deterioration of stainless steel, resulting in corrosion caused by the contact between the primary circuit water environment and the nuclear pressure vessel steel. Therefore, the corrosion behavior of nuclear pressure vessels materials is also summarized in detail. Meanwhile, the evolution mechanism of the microstructure of nuclear pressure vessel materials under thermal aging conditions is analyzed, and the mechanisms affecting the mechanical properties are also described. In addition, the influence mechanisms of internal and external factors on the fatigue properties, fatigue crack initiation and fatigue crack propagation of nuclear pressure vessel steel are analyzed in detail from different perspectives. Finally, the development direction and further research contents of nuclear pressure vessel materials are prospected in order to improve the service life and ensure safe service in harsh environment.
... Lin and Zhang [9,10] conducted a formula for calculating stress concentration and strength of defective cylinder. Zhong [11] studied the fatigue behavior of irradiated reactor pressure vessel steel with experimental method. Their results are very significant, but their formulas are based on certain assumptions and simplifications. ...
Pressure vessel is a primary barrier to gathering system and an important guarantee for normal production in the oilfield. Composite materials for pressure vessels are widely used to adapt to acidic media and temperature and pressure environment. Therefore, the stress corrosion cracking (SCC) is easily caused considering harsh working condition and multi-material structure. Pressure vessel defect due to SCC can cause material strength degradation, medium leakage and even explosion. In order to analyze the stress distribution of the whole equipment and verify the crack detection results in the field, a detailed finite elements model (FEM) of the absorber considering the composite material is established based on mechanical experiments and field data. On the other hand, a WOL specimen FEM is established according to the industry standard. Finally, twelve SCC evaluation experiments of composite materials are carried out based on the results of two established FEM, and the safety windows of 316L and SA516-70 are obtained in specific stress corrosion environments. The work presented in this paper can offer a technological basis and theoretical foundation for strength design of composite materials for pressure vessels in sour environment.
In this study, the low cycle fatigue behavior and failure mechanism of WAAM 16MND5 bainitic steel, which is commonly used in the nuclear reactor, were investigated thoroughly via a series of strain-controlled and stress-controlled low cycle fatigue tests. Results show that the microstructure of WAAM 16MND5 steel is granular bainite with a certain number of banded pro-eutectoid ferrite. WAAM 16MND5 steel exhibits similar cyclic deformation behavior under different strain amplitudes, i.e., a long stage of continuous cyclic softening and rapid cyclic softening till failure. The fatigue life decreases with the increasing strain amplitude. Moreover, the persistent slip markings in ferrite grains, the interface of pro-eutectoid ferrite and large Al2O3 inclusions are the preferred locations for microcrack initiation. The crack initiated from the edge of the specimen is observed to develop into a main crack. The main crack propagates mainly in transgranular mode and can also propagate along the pro-eutectoid ferrite interface.
This paper proposes an alternating bending technique for evaluating fatigue life in the low‐to‐high cycle fatigue regime. A method was developed for estimating the stress, elastic strain, and plastic strain ranges of a plastically deformed specimen subjected to alternating bending with consideration of stress and strain distributions. To evaluate its effectiveness, fatigue testing was conducted using a specimen made of a steel used for pressure vessels. The stress, elastic strain, and plastic strain ranges could be obtained during cyclic bending. The elastic strain amplitude life and plastic strain amplitude life curves were linear in a log–log plot in the low‐to‐high cycle fatigue regime. Hence, the fatigue life under alternating bending could be evaluated using the proposed strain‐based approach. However, these curves could not be predicted using equations with parameters obtained from tensile testing, such as the universal slope method, due to the strain gradient in the specimen.
CMSX-4 is a recently developed rhenium containing single crystal nickel-based superalloy. This alloy has potential applications in many critical high-temperature applications such as turbine blades, rotors, nuclear reactors, etc. The fatigue crack growth rate and the fatigue threshold data of this material is extremely important for accurate life prediction, as well as failure safe design, at elevated temperatures. In this paper, the fatigue crack growth behavior of CMSX-4 has been studied at 650 C. The investigation also examined the influence of [gamma][prime] precipitates (size and distribution) on the near-threshold fatigue crack growth rate and the fatigue threshold. The influence of load ratio on the fatigue crack growth rate and the fatigue threshold was also examined. Detailed fractographic studies were carried out to determine the crack growth mechanism in fatigue in the threshold region. Compact tension specimens were prepared from the single crystal nickel-based superalloy CMSX-4 with [001] orientation as the tensile loading axis direction. These specimens were given three different heat treatments to produce three different [gamma][prime] precipitate sizes and distributions. Fatigue crack growth behavior of these specimens was studied at 650 C in air. The results of the present investigation indicate that the near-threshold fatigue crack growth rate decreases and that the fatigue threshold increases with an increase in the [gamma][prime] precipitate size at 650 C. The fatigue threshold decreased linearly with an increase in load ratio. Fractographs at 650 C show a stage 2 type of crack growth along [l brace]100[r brace] type of crystal planes in the threshold region, and along [l brace]111[r brace] type of crystal planes in the high [Delta]K region.
Local crystallographic configurations (also referred to as local micro-texture) which promote transgranular micro-crack initiation in 316LN stainless steel in low cycle fatigue are studied. Specimens were subjected to tension-compression with constant plastic strain amplitude, in air, at room temperature, during 5000 cycles (i.e. about 20% of the fatigue life). The first part of this work is devoted to a statistical analysis of slip marks and cracks observed at surface of one fatigued specimen using scanning electron microscope (SEM), in a region composed of about 1000 grains. 95 micro-cracks initiated along persistent slip markings detected in this region are analyzed with respect to different characteristics of grains, especially crystallographic orientation, measured using electron backscatter diffraction (EBSD). From the detailed analysis of the numerous data derived from these observations and measurements performed only at surface, the two main significant factors which are found to favour crack formation are the grain size and the orientation of the activated slip system with respect to the surface. Indeed, the mean size of grains which contain cracks is almost twice the one of the remaining grains. Moreover, for most grains in which cracks are observed, the angle between the normal to the surface and the activated Burgers vector (resp. the normal to the activated slip plane) lies in the range [30°, 50°] (resp. [55°, 70°]). No other characteristic was found to provide significant and direct information in order to identify initiation sites. Thus, in the second part of this work, the analysis of initiation sites is performed using additional information relative to three-dimensional (3D) aspects of the microstructure. 3D characterisation of the polycrystalline microstructure and some cracks in one fatigued specimen was achieved using serial-sectioning technique combined with SEM and EBSD. As an example, the study of one specific crack and its surrounding microstructure is presented, including crystal plasticity finite element (CPFE) simulation based on 3D mesh of the polycrystal in this studied region (composed of 386 grains). It is found that the predicted plastic slip activity is more intense within the grain where cracks have been actually observed. This study illustrates that CPFE simulations can provide consistent prediction of slip activity at surface of polycrystals, at least qualitatively, if the actual 3D microstructure is taken into account.
The effect of dynamic strain aging on cyclic stress response and fatigue resistance of ASME SA508 Cl.3 forging steel for nuclear reactor pressure vessels has been evaluated in the temperature range of room temperature to 500°C. Total strain ranges and strain rates were varied from 0.7 to 2.0% and from 4 × 10−4 to 1 × 10−2 s−1, respectively.The cyclic stress response depended on the testing temperature, strain rate, and range. Generally, the initial cyclic hardening was immediately followed by cyclic softening at all strain rates. However, at 300°C, the operating temperature of nuclear reactor pressure vessels, the variation of cyclic stress amplitude showed the primary and secondary hardening stages dependent on the strain rate and strain range. Dynamic strain aging was manifested by enhanced cyclic hardening, distinguished secondary hardening, and negative strain rate sensitivity. A modified cell shutting model was described for the onset of the secondary hardening due to the dynamic strain aging and it was in good agreement with the experimental results.Fatigue life increased in strain rate at all testing temperatures. Specifically the fatigue life was longer at the dynamic strain aging temperature. Further, the dynamic strain aging was easy to initiate the crack, while crack propagation was retarded by crack branching and suppression of plastic zone, hence the dynamic strain aging caused the improvement of fatigue resistance.
Fatigue failure in the high-cycle and ultrahigh-cycle regimes is often dominated by the crack initiation processes, which are strongly influenced by the salient features and defects in the microstructure. Competing fatigue mechanisms involving crack initiation at persistent slipbands, grain boundaries, pores, and non-metallic inclusions or particles, have been reported to occur at surface sites in the high-cycle fatigue regime (106–107 cycles), but shift to interior sites in the ultrahigh-cycle fatigue regime (109–1010 cycles). The changes in fatigue mechanism and crack initiation site result in large variations in fatigue life. This overview article examines the roles of microstructural features such as grain size, texture, porosity, non-metallic inclusion in the fatigue crack initiation process and the manners by which these microstructural effects affect the shape of the stress–life curves.