Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 °C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 °C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 °C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 °C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same reason, the visible dislocation microstructure and radiation-induced hardening are also not sufficient to cause IASCC alone.     

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Isolation of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) was attempted by means of low-dose high-temperature neutron irradiation in a material test reactor to get better understanding on IASCC. Microstructure, grain boundary segregation, hardness and SCC susceptibility were examined on stainless steels irradiated to 0.8 dpa at around 673 K. The irradiation caused well-developed grain boundary segregation without notable hardening or microstructural changes. No IASCC was found in 593 K hydrogenated water whereas the steels were highly susceptible to IASCC in 561 K oxygenated water. The results suggest that grain boundary segregation, probably Cr depletion, is sufficient to cause IASCC in oxygenated water and that other radiation-induced changes such as microstructure and hardening are required for IASCC in hydrogenated water.     

A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.     

Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 °C

Six austenitic stainless steel heats (three heats each of 304SS and 316SS) neutron-irradiated at 275 °C from 0.6 to 13.3 dpa have been carefully characterized by TEM and their hardness measured as a function of dose. The characterization revealed that the microstructure is dominated by a very high density of small Frank loops present in sizes as small as 1 nm and perhaps lower, which could be of both vacancy and interstitial-type. Frank loop density saturated at the lowest doses characterized, whereas the Frank loop size distributions changed with increasing dose from an initially narrow, symmetric shape to a broader, asymmetric shape. Although substantial hardening is caused by the small defects, a simple correlation between hardness changes and density and size of defects does not exist. These results indicate that radiation-induced segregation to the Frank loops could play a role in both defect evolution and hardening response.     

Recovery of Heat-Generating Nuclides in Eluted Solutions

To study about microstructure and chemical composition of oxide films formed on surface of stainless steel is most important for understanding of stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC). In this work, a new sample preparation method for microstructure observation of oxide films was developed. To prevent to break oxide films during fabrication, surface of specimens were protected with plating. Focused ion beam (FIB) processing was conducted to prepare thin foil samples of cross section of oxide films. After sample preparation, microstructure of cross section of oxide films was observed by transmission electron microscope (FE-TEM), and microscopic chemical composition was analyzed by energy dispersed X-ray spectrometer (EDS). From the results, effects of silicon (Si) doping for oxide film formation in two oxidation conditions are discussed.     

Effects of transmutation elements on the defect structure development of W irradiated by protons and neutrons

A series of W-Re-Os alloys were fabricated by arc melting for investigating the effects of transmutation elements of tungsten on the defect structure development. Transmutation electron microscopy has been used to investigate the defect structure for proton-irradiated (E = 1 MeV) W, W-3Re, W-3Os and 0.15 dpa neutron-irradiated (E > 1 MeV) W-5Re-3Os and W, W-3Re, W-5Re and W-26Re. The irradiation-induced voids and dislocation loops which directly cause the irradiation hardening were observed. The results show the combination of W with Re or Os effectively restrains irradiation damage since the number density and radius of both voids and dislocation loops remarkably decrease with increasing Re or Os content.     

A prediction model of IASCC initiation stress for bolts in PWR core internals

A model of IASCC initiation stress for bolts of core internals in pressurized water reactors was developed considering differences in material property changes due to irradiation and material conditions. Assuming that IASCC initiation was controlled by grain boundary composition and yield strength, these values for each specimen of post-irradiation IASCC initiation tests were calculated by physical kinetic models considering dose rate, temperature, material composition and surface hardening. Then, correlations of grain boundary composition and yield strength with IASCC initiation stress were determined. The model predicted that the IASCC initiation stress became lower with dose and was lower for higher temperature, lower flux and higher surface hardening level.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

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.     

Mechanism of irradiation assisted stress corrosion crack initiation in thermally sensitized 304 stainless steel

Thermally sensitized 304 stainless steels, irradiated up to 1.2 × 10²¹ n/cm² (E > 1 MeV), were slow-strain-rate-tensile tested in 290 °C water containing 0.2 ppm dissolved oxygen (DO), followed by scanning and transmission electron microscopic examinations, to study mechanism of irradiation-assisted-stress-corrosion-crack (IASCC) initiation. Intergranular (IG) cracking behaviors changed at a border fluence (around 1 × 10²⁰ n/cm²), above which deformation twinning were predominant and deformation localization occurred earlier with increasing fluence. The crack initiation sites tended to link to the deformation bands, indicating that the crack initiation may be brought about by the deformation bands interacted with grain boundaries. Thus the border fluence is equivalent to the IASCC threshold fluence for the sensitized material, although the terminology of IASCC is originally given to the non-sensitized materials without microstructural definition. The IASCC threshold fluence was found to change with irradiation conditions. Changes in IASCC susceptibility and IASCC threshold fluence with fluence and DO were further discussed.     

纳米压痕技术研究堆内构件材料304NG不锈钢的三束辐照效应

在中国原子能科学研究院的三束辐照实验平台上,对国产堆内构件材料核级控氮304NG不锈钢进行重离子、氢和氦同时辐照,并采用连续刚度纳米压痕技术测量了辐照前后样品的硬度和弹性模量的变化。结果表明,6dpa下,随着辐照温度的升高,辐照硬化减弱;300℃下,随着辐照剂量的增加,辐照硬化增大,高剂量情况下,氢、氦注入区的辐照硬化更显著,表明存在氢、氦增强硬化效应。     

In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 10²⁵ n/m² under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specimens, the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate was considered to be small, because the in-core data under the same ECP condition were similar to the ex-core data under the DO = 32 ppm condition.     

Impact of localized deformation on IASCC in austenitic stainless steels

Localized deformation has been identified as a potential primary contributor to IASCC. Seven austenitic alloys were irradiated to 1 and 5 dpa at 360 °C using 2-3.2 MeV protons and were tested both in simulated BWR environment and in argon. Cracking susceptibility was evaluated at both 1% and 3% strain intervals using crack length per unit area. Stacking fault energy (SFE), hardness, radiation-induced segregation (RIS) and localized deformation were characterized and their correlations with cracking were evaluated using a proposed term, correlation strength. Both SFE and hardness contributed to cracking but neither was the dominant factor. RIS did not play an important role in this study. The correlation strength of localized deformation with IASCC was found to be significantly higher than for others parameters, implying that localized deformation is the most important factor in IASCC. Although not well understood, localized deformation may promote cracking through intensive interaction of dislocations in slip channels with grain boundaries.     

Materials aspects of BWR plant life extension

Considerable experience with plant equipment performance in nuclear power stations has indicated that the principal factors limiting the life of BWRs and PWRs are materials related. Specifically, for LWRs it is known that these materials issues generally include parameters related to stress corrosion cracking, corrosion fatigue, wear and radiation embrittlement. Not only do these parameters affect and limit the actual useful design life of plant components but also affect the plant's operating availability. In all these cases, the elimination or control of one or more of these critical parameters should improve the plants availability and significantly extend the useful service life.In the present paper, research performed to address the intergranular stress corrosion cracking (IGSCC) area is described. Specific emphasis is placed on Type 304 stainless steel which has suffered IGSCC in piping in the heat-affected-zone (HAZs) adjacent to the welds in the BWR primary system. Research has developed and qualified a number of techniques which address the three necessary conditions for IGSCC in BWRs: (1) sensitized microstructure, i.e., chromium depletion at the grain boundaries during welding; (2) over yield tensile stress; and (3) oxygenated (200 ppb) high temperature (288Another potential life-limiting IGSCC phenomenon for certain components, irradiation assisted stress corrosion cracking (IASCC) of stainless steel exposed to a high neutron flux, is also discussed. Unlike the IGSCC, IASCC results in intergranular cracking of annealed material at low stress. Fortunately, preliminary research has indicated that some of the techniques utilized for IGSCC control in piping as well as new controlled impurity level stainless steel alloys may reduce the future potential IASCC concern to an insignificant level.     

Fatigue Crack Initiation in Proton-Irradiated Austenitic Stainless Steel

The effect of irradiation on slip band formation and growth and microcrack initiation behavior under low cycle fatigue in SUS316L austenitic stainless steel was investigated using accelerator-based proton irradiation and a low cycle fatigue test at room temperature in air. The mean space of the slip line in proton-irradiated specimens was 25–40% wider than that in unirradiated specimens under the same number of cycles, possibly due to localized deformation by proton irradiation. The microcrack initiation life of the proton-irradiated specimens was approximately 20% of that of the unirradiated specimens. While the microcrack initiation in the unirradiated specimens was observed at the grain boundary, twin boundary, slip band, and triple junction, that in the proton-irradiated specimens was observed only at the twin boundary and slip band, possibly due to irradiation hardening. The step-height of an extrusion near the microcrack was almost the same in the unirradiated and proton-irradiated specimens regardless of the initiation site (100–150 nm). Therefore, the microcrack initiation was considered to occur when the surface morphology change involving the extrusion exceeded the specific threshold value.     

The effect of prior cold-work on the deformation behaviour of neutron irradiated AISI 304 austenitic stainless steel

Cold-work is intentionally employed to increase the yield strength of austenitic stainless steels and also occurs during fabrication processes, but it has also been associated with greater incidence of stress corrosion cracking. This study examined the effect of up to 3.85 dpa neutron irradiation on the deformation behaviour and microstructures of 30% cold-worked AISI 304 material tensile tested at 300 °C. While the deformation behaviour of 0.07 dpa material was similar to non-irradiated material tested at the same temperature, its stress-strain curve was shifted upwards by about 200 MPa. Materials irradiated to over 2 dpa hardened some 400-500 MPa, but showed limited strain hardening capacity, exhibiting precipitous softening with further straining beyond the yield point. The observed behaviour is most likely a consequence of planar deformation products serving as strengtheners to the unirradiated bulk on the one hand, while promoting strain localization on the other, behaviour exacerbated by the subsequent neutron irradiation.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Hardening and Recovery of Neutron-Irradiated Fe-Cu-C Alloy

Internal friction measurements were performed on neutron-irradiated and annealed Fe-0.2w/0Cu-0.0066w/0C alloy, in an attempt to gain information on the role of interstitial C atoms in radiation effects. It was revealed that C atoms are trapped by defects produced by neutron-irradiation, and released upon annealing at about 400°C. The radiation hardening induced in the neutron-irradiated Fe-Cu-C alloy recovers by annealing in at least three steps of 150°–250°C, 350°–450°C and above 550°C. The second step corresponds to the release of C atoms from traps.     

Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part I: Experimental results

Solution annealed 304L (SA 304L) and cold work 316 (CW 316) austenitic stainless steel irradiation creep behaviour have been studied thoroughly. Irradiations were carried out in fast breeder reactors BOR-60 (at 330 °C, up to 120 dpa) and EBR-II (at 375 °C, up to 10.5 dpa), and in the OSIRIS mixed spectrum reactor (at 330 °C, up to 9.8 dpa). After an incubation threshold, the irradiation creep of the austenitic stainless steels is linear in stress and in dose. Creep appears to be athermal in this temperature range. A significant difference in the behaviour is measured between the creep of SA 304L and CW 316.In order to study the anisotropy of loop population, which would be the signature of a possible stress induced preferential absorption (SIPA) mechanism for irradiation creep, special attention was given to the measurement of anisotropy of loop distribution between the four families. The anisotropy induced by an applied stress has been shown to be in the range of the statistical scatter in the situation where no stress is applied. TEM microstructural analyses performed on this sample show slight difference between the microstructure of specimens deformed under irradiation and the microstructure of specimens irradiated without stress under the same irradiation conditions.     

ODS-Eurofer钢微观结构及辐照硬化研究

研究了ODS-Eurofer钢的微观结构及辐照硬化现象。首先用透射电子显微镜(TEM)观察了ODS-Eurofer钢的初始微观组织结构,发现基体中不仅存在几nm至几十nm的氧化物弥散颗粒,还存在具有壳 核结构的大尺寸(直径大于100 nm)颗粒,并观察到纳米颗粒对位错线的钉扎作用。随后用能量为5 MeV的Fe²⁺离子在300 ℃和500 ℃下辐照样品至25 dpa以模拟中子辐照,并用纳米压痕仪和TEM测试表征了辐照所致力学性能和微观结构的变化。结果表明,两种温度下辐照均引起硬度上升,500 ℃时由于辐照产生的点缺陷发生复合,导致硬化效应弱于300 ℃。用TEM观测辐照水平为25 dpa的损伤层发现有少量纳米尺寸位错环,这些位错环是辐照硬化的主要原因。ODS-Eurofer钢初始微观结构对辐照硬化有重要影响,其中晶界、纳米颗粒与基体界面、位错线等能捕获辐照过程中产生的点缺陷,从而抑制辐照位错环的生长。