Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

Cracking in stabilized austenitic stainless steel piping of German boiling water reactors—characteristic features and root cause

Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur.     

Precipitates induced in austenitic stainless steel by nitrogen injection

Samples of solution-annealed Type 316 stainless steel were irradiated at 803 K with 1.1 MeV nitrogen ions up to 60 dpa at peak and the precipitates induced by the irradiation were examined by electron microscopy. The ratio of injected nitrogen concentration to dpa is estimated to be about 0.3 at%/dpa in an observed area. An ordered γ'-phase of nominally Ni₃Si is identified to form in the sample irradiated to 20 dpa; it shows superlattice spots in the electron diffraction pattern and enrichment of Si and Ni at the precipitate in EDS analysis. In the sample irradiated to 42 dpa, platelet precipitates of 18 nm in average diameter were observed with the number density of 7.4 × 10²¹/m³. The precipitate grows to an average diameter of 28 nm with little change in number density during irradiation from 42 to 60 dpa. The precipitate formed in the sample irradiated to 60 dpa was confirmed to be in the CrN phase with electron diffraction patterns and the spacing of moiré fringes.     

Radiation-induced segregation in desensitized type 304 austenitic stainless steel

Radiation-induced segregation (RIS) in desensitized type 304 stainless steel (SS) was investigated using a combination of electrochemical potentiokinetic reactivation (EPR) test and atomic force microscopy (AFM). Desensitized type 304 SS was irradiated to 0.43 dpa (displacement per atom) using 4.8 MeV protons at 300 °C. The maximum attack in the EPR test for the irradiated desensitized SS was measured at a depth of 70 μm from the surface. Grain boundaries and twin boundaries got attacked and pit-like features within the grains were observed after the EPR test at the depth of 70 μm. The depth of attack, as measured by AFM, was higher at grain boundaries and pit-like features as compared to twin boundaries. It has been shown that the chromium depletion due to RIS takes place at the carbide-matrix as well as at the carbide-carbide interfaces at grain boundaries. The width of attack at grain boundaries after the EPR test of the irradiated desensitized specimen appeared larger due to the dislodgement of carbides at grain boundaries.     

Characterization of 316 stainless steel surfaces used in BWR recirculation piping

The effect of surface treatments such as sandblasting, acid pickling, and electropolishing on surface topography and composition has been investigated for 316 nuclear grade stainless steel. The build-up of radioactive cobalt-60 on sandblasted surfaces is much higher compared to that on acid pickled surfaces. Compared to sandblasted surfaces, acid pickling produces smooth surfaces and also generates grain boundary grooves, which can be effectively removed by subsequent electropolishing. Sandblasting produces rough and textured surfaces. Electropolishing of such surfaces reduces the roughness but does not produce as smooth a surface as a pickled surface. The techniques used in the surface analyses were scanning electron microscopy (SEM), X-ray energy spectroscopy (XES), and Auger electron spectroscopy (AES).     

Serrated flow behavior in AL6XN austenitic stainless steel

Serrated flow behavior of the AL6XN austenitic stainless steel has been investigated at different temperatures and strain rates. The results show the serrated flow, peak/plateau in flow stress and negative strain rate sensitivity appearing in tensile deformation of the AL6XN steel at 773-973 K and 3.3 × 10⁻⁵-3.3 × 10⁻³ s⁻¹ (excluding 873 K, 3.3 × 10⁻⁵ s⁻¹), suggesting the occurrence of dynamic strain aging (DSA). The activation energy for type-A and -(A + B) serrations was calculated to be 304 kJ/mol and diffusion of substitutional solutes, such as chromium and molybdenum is considered as the mechanism of serrated flow. TEM observations further revealed a typical planar slip mode in the regime of DSA of the deformed AL6XN steel.     

Mechanisms of stress relief cracking in titanium stabilised austenitic stainless steel

The heat affected zone (HAZ) of AISI 321 welds may exhibit a serious form of cracking during service at high temperature. This form of damage, called ‘stress relief cracking’, is known to be due to work hardening but also to aging due to Ti(C,N) precipitation on dislocations which modifies the mechanical behaviour of the HAZ. The present study aims to analyse the latter embrittlement mechanism in one specific heat of 321 stainless steel. To this end, different HAZs are simulated using an annealing heat-treatment, followed by various cold rolling and aging conditions. Then, we study the effects of work hardening and aging on Ti(C,N) precipitation, on the mechanical (hardness, tensile and creep) behaviour of the simulated HAZs and on their sensitivity to intergranular crack propagation through stress relaxation tests performed on pre-cracked CT type specimens tested at 600 °C. It is shown that work hardening is the main parameter of the involved mechanism but that aging does not promote crack initiation although it leads to titanium carbide precipitation. Therefore, the role of Ti(C,N) precipitation on stress relief cracking mechanisms is discussed. An attempt is made to show that solute drag effects are mainly responsible for this form of intergranular damage, rather than Ti(C,N) precipitation.     

Study of low-cycle fatigue of AL6XN austenitic stainless steel

The low-cycle fatigue (LCF) behaviors of the AL6XN austenitic stainless steel were investigated at strain rates of 3.3 × 10⁻⁵ to 3.3 × 10⁻³ s⁻¹ and at different temperatures, respectively. The weakened cyclic softening, negative strain-rate-stress response and anomalous temperature-dependence of cycle stress reflect the dynamic strain aging (DSA) hardening during LCF at elevated temperatures. Electron microscopy observations revealed that the dislocation structure changes from the cellular structure at room temperature to the planar slip band, serving as crack initiation sites in the regime of DSA. The DSA hardening results in the reduction of fatigue resistance at elevated temperature via reducing the crack initiation and propagation life.     

钼离子注入奥氏体不锈钢引发亚结构的变化

用加速电压为５０ｋＶ，剂量为２×１０＾１７ｉｏｎｓ／ｃｍ＾２的钼离子注入经１１５０℃固溶处理的奥氏体不锈钢中，利用透射电子显微镜弱束成像技术研究了被注入材料的横截面样品，发现离子辐照引发被注入亚结构变化的深度远大于离子本身的注入深度，其亚结构为多种位错组成的复杂位错组态，这种亚结构会使被注入材料具发于加工硬化的效果。探讨了离子注入材料中由辐射引发的加工硬化现象，即所谓的“长程效应”的原因。     

High velocity tensile test of austenitic stainless steel at elevated temperatures

This paper presents experimental results concerning the tensile properties of JIS Type SUS 316 stainless steel. The test was carried out at room temperature, 400°C and 550°C at strain rates of 10⁻³ 1/s and 10² 1/s. Base metal, weld joint and weld metal specimens were chosen for the test. The aim of this test is to clarify the effects of strain rate and test temperature on the mechanical properties such as 0.2% yield strength, ultimate tensile strength and elongation of JIS Type SUS 316 stainless steel.     

Void swelling and precipitation in a titanium-modified austenitic stainless steel under proton irradiation

The correlation between void swelling and precipitation behavior in a 10% cold worked Fe-16.2Ni-14.6Cr-2.37Mo-1.79Mn-0.53Si-0.24Ti-0.06C alloy was examined with 200 keV proton irradiation. Swelling peak temperature after the proton irradiation to 10 dpa was about 823 K, and void swelling decreased steeply with increase in irradiation temperature from 823 to 923 K. Void swelling increased rapidly from 1.9 to 12.1% with increase in irradiation dose from 20 to 45 dpa at 873 K. Fine intragranular TiC precipitates, which were formed during initial stage of irradiation, dissolved gradually with increase in irradiation dose from 10 to 45 dpa at 873 K, while the amount of precipitation of needle-shaped Fe₂P phase containing titanium increased with increasing dose. The reduction of sink strength of the TiC precipitates due to the dissolution during irradiation was thought to cause the increase of swelling rate with increase in irradiation dose from 20 to 45 dpa at 873 K.     

Development of inelastic constitutive model for austenitic stainless steel for design use

To prevent creep-fatigue failure or excessive deformation in high-temperature components of fast reactor plants, accurate estimation of inelastic deformation is essential. In performing inelastic analysis, employment of constitutive models, which can precisely reproduce inelastic deformation of the material is of critical importance. The authors have been engaged in the development of inelastic constitutive model for the use in structural design assessment of liquid metal-cooled fast reactor plants. Various improvements were made on the nonlinear hardening model proposed by Ohno and Wang, placing an emphasis on capability to simulate inelastic deformation behavior of austenitic stainless steels, under regular or irregular cyclic loading possibly with temperature variation and hold time. It was demonstrated that the model can simulate the inelastic deformation behavior under various loading conditions with a sufficient accuracy.     

Modelling the high temperature mechanical behaviour of helium embrittled austenitic stainless steel

Current theories for the growth of grain-boundary defects during plastic deformation have been used to model the ductility of an irradiated austenitic steel as a function of strain rate and thermal-neutron fluence. Failure of irradiated steels is via the growth and linkage of helium bubbles on the grain boundaries produced by an (n, ) reaction between the thermal neutrons and boron present in the steel as a controlled trace element.

Reasonable agreement is obtained between theoretically predicted and experimentally measured ductilities over a range of strain rates between 10⁻¹⁰ and 10⁻²/s. However, in order to obtain such agreement over the range of strain rates examined, variables such as defect size and spacing had to be carefully selected. The values used were internally self-consistent with the amount of helium known to be availablc to nucleate grain-boundary bubbles.

The models of defect growth have also been used to make some predictions on the creep-fatigue performance of such materials.     

Ductile fracture behavior of circumferentially cracked type 304 stainless steel piping under bending load

A ductile pipe fracture test program has been conducted in Japan Atomic Energy Research Institute (JAERI) to investigate the ductile fracture behavior of circumferentially cracked pipes and to demonstrate the validity of the leak before break concept in LWR pipings.In the paper are described the scope of the pipe test program and current test results for 6-inch diameter type 304 stainless steel pipes. Test pipes with a through-wall or a part-through crack in the circumferential direction were bent under low or high compliance condition, and stable or unstable pipe fracture behavior was investigated. J based tearing instability criterion and the net section collapse criterion are compared with the pipe test results, and the validity of these fracture criteria is discussed. Furthermore, geometries of acceptable flaws in pipes are evaluated considering the leak before break condition.     

Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO₂ laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 °C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.     

Evolution of cavity size distributions in dual-ion irradiated austenitic stainless steel and Fe-Cr-Ni ternary alloys

The first four moments of experimentally measured cavity size distributions in dual-ion irradiated 304SS, Fe-12Ni-15Cr and Fe-30Ni-15Cr alloys have been calculated for a range of fluences, helium injection rates, and irradiation temperatures. The moments are shown to correctly describe the effects of alloy composition, fluence, helium and temperature on the evolution of the cavity size distributions. Experimentally determined moments are compared with those calculated from cavity nucleation and growth theories. The moments reflect the competition between nucleation and growth processes and provide insight into the details of the transient low swelling regime.