Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water

A formulation for the quantitative calculation of the stress corrosion cracking (SCC) growth rate was proposed based on a fundamental-based crack tip strain rate (CTSR) equation that was derived from the time-based mathematical derivation of a continuum mechanics equation. The CTSR equation includes an uncertain parameter r₀, the characteristic distance away from a growing crack tip, at which a representative strain rate should be defined. In this research, slow strain rate tensile tests on sensitized 304L stainless steel in oxygenated high temperature water were performed. By curve fitting the experimental results to the numerically calculated crack growth rate, the parameter r₀ was determined. Then, the theoretical formulation was used to predict the SCC growth rates. The results indicate that r₀ is on the order of several micrometers, and that the application of the theoretical equation in predicting the crack growth rate provides satisfactory agreement with the available data.     

Preliminary results of post-irradiation examinations on LiSoR-2 test section

Liquid-solid reaction under irradiation (LiSoR) experiments are aimed at understanding the effects of liquid lead-bismuth eutectic (LBE) corrosion and embrittlement under irradiation on structural materials, which is one of the key items of the materials R&D for the future accelerator-driven system (ADS). The LiSoR setup is basically a LBE loop with a test section irradiated with 72 MeV protons. The second irradiation was conducted for about 34 h and terminated after a leakage of LBE was detected. Post-irradiation examinations (PIE) are being performed on both the tube and tensile specimen in the test section. Optical microscopy, scanning electron microscopy, transmission electron microscopy and microhardness tests have been completed. The results show that a crack formed in the irradiation zone of the tube. In the material in the irradiation zones of both the tube and the tensile specimen dislocation cell structure is well developed, which indicates heavy deformation due to thermal fatigue. The crack should start at the inner surface and propagate to the outer surface. The fracture surfaces of the crack are dominated by a brittle cleavage fracture mode. However, on the surfaces of the tensile specimen, no microcracks are observed.     

Tensile specimen geometry and the constitutive behavior of Zircaloy-4

The influence of tensile specimen geometry on the deformation behavior of flat Zircaloy-4 tensile specimens has been examined for gauge length-to-width ratios that range from 1:1 to 4:1. Specimen geometry has only minor effects on the values of the yield stress, tensile strength, apparent uniform strain at maximum load, and strain-hardening exponent. However, in all geometries but the 4:1 configuration, diffuse necking occurs before maximum load. As a result, strain distributions at maximum load are uniform only in the 4:1 geometry. The elongation to failure is also affected by specimen geometry with the shorter gauge sections exhibiting much higher total elongation values, due in large part to the concomitant specimen necking behavior.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10¹⁸ and 1.02 × 10¹⁹ n/cm² (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.     

Effect of prior thermal treatment on the microchemistry and crack propagation of proton-irradiated AISI 304 stainless steels

The effect of prior thermal treatment on crack growth was investigated on proton-irradiated Type 304 stainless steel (SS) of initially solution annealed (SA) and thermally sensitized (SEN) conditions. The Cr depletion profiles were measured by field emission gun transmission electron microscopy/energy dispersive spectroscopy (FEGTEM/EDS) in an attempt to correlate grain boundary chromium composition with the measured crack growth rate. The results showed that the crack growth of the 1-dpa-irradiated SEN 304SS is substantially higher than that of SA 304SS with the same irradiation dose. The unirradiated SEN material initially started with a shallow Cr depletion profile near grain boundary. After 1 dpa irradiation with proton, the Cr depletion profile becomes narrower and deeper. In contrast, the grain boundary Cr concentration in the SA specimen at the same irradiation dose was higher than that of the SEN specimen, mainly due to an initial Cr enriched condition. Consequently, the irradiated SEN specimen exhibited a higher degree of sensitization in electrochemical potentiokinetic reactivation test and faster crack growth rate in the stress corrosion crack test. The absence of irradiation enhanced crack growth in heavily thermal-sensitized 304SS is probably attributed to slower radiation-induced Cr depletion as a result of pre-existing thermally induced grain boundary Cr depletion. It is a clear indication that the inverse Kirkendall effect was hampered by the back diffusion of Cr due to initially depleted Cr concentration gradient near grain boundary.     

Failure behavior of Zircaloy-4 cladding after oxidation and water quench

Simulated LOCA (loss of coolant accident) tests and subsequent mechanical tests on Zircaloy-4 cladding were carried out to evaluate the failure behavior of the cladding. Zircaloy-4 claddings were oxidized in a steam environment from 900 to 1250 °C for a given time period followed by a flooding of cool water to simulate LOCA tests. After the simulated LOCA test, the ductility of the oxidized cladding was evaluated by mechanical tests such as ring compression test and 3-point bend test. Evaluation of the absorbed contents such as hydrogen and oxygen were also carried out. The results showed that Zircaloy-4 cladding failed during thermal shock when the ECR (equivalent cladding reacted) value exceeded 20%. Lower boundary of brittle failure at thermal shock corresponds to 20% of ECR line calculated by the Baker-Just equation regardless of test temperature. On the other hand, boundary of ductile failure by the mechanical test did not followed after the ECR line. It rapidly decreased above 1000 °C to show that all Zircaloy-4 claddings behaved brittle fracture above 1150 °C when it oxidized at 300 s. Microstructural analysis revealed that boundary of ductile failure by the mechanical test fitted well when the absorbed oxygen content inside the prior-β layer was below 0.5 wt%.     

辐照用小尺寸样品力学性能表征技术研究进展

由于辐照空间尺寸限制、降低样品放射性和提高辐照参数精度等原因，小尺寸样品被广泛应用于核反应堆材料的辐照后力学性能表征。本文就国内外小尺寸拉伸、冲击、断裂韧性、疲劳、蠕变和小冲杆等测试表征技术的研究现状进行了综合论述，分析了小尺寸样品测试中的关键影响因素以及数据归一化方法，总结了小尺寸样品存在的问题，并结合我国需求对小尺寸样品技术的发展进行了分析和展望，以期为小尺寸样品技术及测试分析数据进一步规范化和工程应用发展提供参考。     

Effects of natural aging on the tensile properties of water-quenched U-6%Nb alloy

Uranium-6 wt% niobium (U-6%Nb) alloy has been in use for many years in the water-quenched (WQ) condition. The purpose of this work was to determine the effect of natural aging on tensile properties of the WQ U-6%Nb alloy. The materials studied were hemispherical shells after 15 and 20 years in storage. The alloy was successfully tested in the original curved configuration, using the specially designed tensile test apparatus. Finite element analysis confirmed the validity of the test method. The results of the tensile tests clearly indicated that in the WQ condition, the material is changing and after 15 and 20 years, the yield strength exceeds the original maximum allowable specification. The fracture mode transitions from highly ductile, microvoid coalescence in new material to a mixed mode of shallow dimples and inclusion-induced voids in the naturally aged material.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

Evaluation of ductile–brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

Small punch (SP) tests were performed to evaluate the ductile–brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm × 10 mm × 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 °C. Small punch tests were performed in the temperature range of RT to −196 °C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T_SP), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T_41 J. T_SP from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test (ΔT_41 J) shows a better correlation with the transition temperature shift (ΔT_SP(E)) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T_SP also had a correlation with the reference temperature (T₀) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.     

Liquid metal embrittlement of T91 martensitic steel evidenced by small punch test

In this work, the sensitivity of liquid metal embrittlement of the T91 martensitic steel is investigated with the small punch test (SPT). The material was studied in three tempering conditions (as quenched, tempered at 500 and 750 °C), at 300 °C in air and in the liquid lead bismuth eutectic (LBE). The load–displacement curves (four stages, low maximum force and large displacement to fracture) obtained for one test condition of the 750 °C tempered material is in general very different from those of the two other materials. An effect of LBE has been observed for the as quenched and 500 °C tempered steels. For these materials, the curves tend to be linear with a reduced displacement to fracture suggesting a brittle behavior. This ductile to brittle transition induced by liquid metal has been confirmed from the fracture surface analysis where cleavage was observed. In comparison with conventional tensile tests, small punch tests appear to be more sensitive to evidence liquid metal embrittlement.     

Critical strains and necking phenomena for different steel sheet specimens under uniaxial loading

When defining ultimate loads or failure criteria for shells, vessels and containment liners, for instance, reference will be made to critical strains of the material under consideration.In general the critical strains are derived from uniaxial tensile test results obtained with rods.The actual deformation behaviour of sheets (plate and shell type structures), however, differs considerably from the uniaxial rod behaviour.From various tests - along with theoretical investigations - it was found that for membrane or even bending stressed sheets higher ultimate strains are reached.The typical strain behaviour caused by increasing load starts with uniform strain distribution over the specimen length and finally ends with localized necking.As “critical strains” the well known uniform elongation strain and the ultimate strain are defined.For sheets in addition the quasi uniform elongation strain has been introduced in this paper.Whereas for rods under tension localized necking is directly following the uniform straining, tests with sheets show a so-called diffuse necking behaviour before localized necking starts.As a consequence the deformation capacity of steel sheets turns out to be significantly higher than relying on the data from standard tension tests with rods.     

Corrosion behaviors of US steels in flowing lead-bismuth eutectic (LBE)

Corrosion tests of several US martensitic and austenitic steels were performed in a forced circulation lead-bismuth eutectic non-isothermal loop at the Institute of Physics and Power Engineering (IPPE), Russia. Tube and rod specimens of austenitic steels 316/316L, D-9, and martensitic steels HT-9, T-410 were inserted in the loop. Experiments were carried out simultaneously at 460 °C and 550 °C for 1000, 2000 and 3000 h. The flow velocity at the test sections was 1.9 m/s and the oxygen concentration in LBE was in the range of 0.03-0.05 wppm. The results showed that at 460 °C, all the test steels have satisfactory corrosion resistance: a thin protective oxide layer formed on the steel surfaces and no observable dissolution of steel components occurred. At 550 °C, rod specimens suffered rather severe local liquid metal corrosion and slot corrosion; while tube specimens were subject to oxidation and formed double-layer oxide films that can be roughly described as a porous Fe₃O₄ outer layer over a chrome-rich spinel inner layer. Neglecting the mass transfer corrosion effects by the flowing LBE, calculations based on Wagner’s theory reproduce the experimental results on the oxide thickness, indicating that the oxide growth mechanism of steels in LBE is similar to that of steels in air/steam, with slight modification by dissolution and oxide dissociation at the liquid metal interface.     

Determination of neutron exposure of AISI 304 stainless steel from a BWR top guide using retrospective dosimetry

Retrospective dosimetry was used to determine the accumulated neutron exposure of AISI 304 stainless steel removed from the top guide of a boiling water reactor located at the Oyster Creek nuclear power station. The material was removed from areas adjacent to cracks that were observed after ∼20 years of operation. Using the plant operational history and a variety of measurements of various radioisotopes or non-radioactive transmutation products produced by irradiation, it was possible to determine the integrated neutron fluence experienced by the cracked region and to specify the accumulated displacement dose. Dose estimates on two separate specimens adjacent to the cracks were found to average 1.5 ± 0.2 dpa, possibly reflecting some uncertainty in measurement but more likely suggesting a small gradient in neutron flux-spectra within the section from which the various analysis specimens were cut. This report demonstrates that it is possible to examine defective components lying outside of the core region and where neutron flux-spectra are not well known, and to use the induced transmutation products to determine the neutron exposure with some confidence by using the examined specimen as its own dosimeter.     

Dissimilar welding of nickel-based Alloy 690 to SUS 304L with Ti addition

This study investigates the effects of Ti addition on the weldability, microstructure and mechanical properties of a dissimilar weldment of Alloy 690 and SUS 304L. Shielding metal arc welding (SMAW) is employed to butt-weld two plates with three welding layers, where each layer is deposited in a single pass. To investigate the effects of Ti addition, the flux coatings of the electrodes used in the welding process are modified by varying additions of either a Ti-Fe compound or a Ti powder. The results indicate that the microstructure of the fusion zone (FZ) is primarily dendritic. With increasing Ti content, it is noted that the microstructure changes from a columnar dendritic to an equiaxed dendritic, in which the primary dendrite arm spacing (PDAS) becomes shorter. Furthermore, it is observed that the amount of Al-Ti oxide phase increases in the inter-dendritic region, while the amount of Nb-rich phase decreases. Moreover, the average hardness of the FZ increases slightly. The results indicate that Ti addition prompts a significant increase in the elongation of the weldment (i.e. 36.5%, Ti: 0.41 wt%), although the tensile strength remains relatively unchanged. However, at an increased Ti content of 0.91 wt%, an obvious reduction in the tensile strength is noted, which can be attributed to a general reduction in the weldability of the joint.     

Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

Uniaxial true stress-true strain curve for a ductile metal

It is well known and has been shown qualitatively and quantitatively that the necking phenomenon in uniaxial tensile ductile metal specimens is caused by the presence of non-uniformities in the initial geometry and initial strength. It is shown in this paper that a uniaxial true stress-true strain curve for an idealized perfect specimen could be constructed from experimental data which is available for uniaxial tensile tests on ductile metal specimens having a known initial non-uniformity.     

Embrittlement of RAFM EUROFER97 by implanted hydrogen

About 380 μm thick specimens of low-activation martensitic stainless steel EUROFER97 were homogeneously implanted with protons below about 70 °C to concentrations up to about 1200 appm. Tensile tests were performed at 25 and 200 °C. The tests at 25 °C showed an increase of yield stress and ultimate tensile strength and a decrease of uniform elongation and elongation to fracture, while effects at 200 °C were strongly reduced. Scanning electron microscopy showed virtually no change of the ductile, transgranular fracture mode by the implanted H. A slight decrease of necking was observed only at the highest concentrations. The results are compared to previous measurements on F82H-mod and to literature results on tensile tests after He implantation and neutron irradiation. F82H-mod specimens were also implanted under applied tensile stress to concentrations up to 1900 appm. Straining was ascribed to accumulation of atomic defects, but no fracturing occurred. Thermal desorption measurements are included on H-implanted and tensile tested F82H-mod specimens. The results show significant data scatter, even within one specimen, with the highest measured values being in accordance with the implanted amount. In general, the H content is decreased after testing at higher temperatures, but even after testing at 350 °C, some specimens contain significant amounts.     

国产A508-3钢小尺寸拉伸样品的拉伸颈缩行为研究

对不同厚度国产A508-3钢小尺寸拉伸样品进行了室温拉伸试验,分析了拉伸性能及颈缩段参数,并基于有限元逆运算构建了小尺寸拉伸样品拉伸过程的GTN（Gurson-Tvergaard-Needleman）细观损伤模型,研究了厚度对小尺寸拉伸样品拉伸颈缩行为的影响规律与机理。试验结果表明,小尺寸拉伸样品在变形过程中发生了弹性变形、均匀塑性变形和颈缩变形;随着样品厚度由0.75 mm降低至0.30 mm,屈服强度、抗拉强度和均匀延伸率无明显变化,非均匀延伸率及总延伸率逐渐降低,颈缩角逐渐增大,断裂角在厚度降低至0.50 mm后逐渐增大。GTN细观损伤模型中用于表征空洞形核和融合率的参数在0.30 mm样品中明显降低,此结果与小尺寸拉伸样品颈缩行为规律相互印证。