国产快堆包壳材料CW316(Ti)SS高温强度下降的微观机理分析

冷加工316(Ti)不锈钢CW 316(Ti)SS是我国首选的快堆包壳材料,国产材料的常规力学性能与国外数据相当,但高温蠕变和高温持久强度数据却较低.本项研究主要是通过观察、比较国产快堆包壳材料和俄罗斯快堆包壳材料在高温下微观结构的变化情况,并结合对国产材料高温持久断裂试验样品的断口形貌观察结果,分析得出:国产材料长时高温力学性能下降的主要原因是沿晶界的σ相析出.     

Emulation of neutron irradiation effects with protons: validation of principle

This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 °C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 °C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 °C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a ‘W' shaped profile at 1.0 dpa and then into a ‘V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 °C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy.     

Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10²⁵ neutrons.m⁻², E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10²³ m⁻³) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.     

Influence of helium injection schedule and prior thermomechanical treatment on the microstructure of type 316 SS

The influence of different helium injection schedules on microstructure development in Ni⁺ ion-irradiated 316 SS at 625°C is discussed. Injection schedules were chosen to either approximate the MFR condition or mimic the mixed-spectrum reactor condition. Dual-ion irradiation to 25 dpa produced strongly bimodal cavity size distributions in solution annealed and solution annealed and aged samples, whereas single-ion irradiation followed by dual-ion irradiation to the same dose produced a cavity size distribution with a substantial component of intermediate-size cavities. Dual-ion irradiation produced only very small cavities in 20% CW material, while single-ion followed by dual-ion irradiation produced some intermediate size cavities and greater swelling.     

Temperature and Dose Dependence of Swelling in 10% and 20% Cold-Worked Type 316 Stainless Steels

Void swelling in 10% cold-worked (10% CW) and 20% cold-worked (20% CW) type 316 stainless steels was investigated by 200 keV C⁺ ion irradiation and transmission electron microscope observation. Both 10% CW and 20% CW 316 steels show the swelling maximum at 923 K. Swelling in 10% CW 316 is much higher than that in 20% CW 316. The voids in the former material are larger and fewer than those in the latter material. The bilinear equation is applicable to describe swelling dose relation for both materials, except 10% CW 3 16 at higher doses than 50 dpa, where sharp swelling increase is observed. Heat to heat variability seems to exist in incubation dose, though it is not large. With regard to swelling rate, all three heats examined show good coincidence for both 10% CW and 20% CW 316 steels. Comparison of 20% CW 316 swelling rate for various irradiation projectiles indicates that the swelling rate is described as a simple function of the projectile mass, and there may exist a scaling law between the different projectile data.     

Proton irradiation emulation of PWR neutron damage microstructures in solution annealed 304 and cold-worked 316 stainless steels

Solution annealed (SA) 304 and cold-worked (CW) 316 austenitic stainless steels were pre-implanted with helium and were irradiated with protons in order to study the potential effects of helium, irradiation dose, and irradiation temperature on microstructural evolution, especially void swelling, with relevance to the behavior of austenitic core internals in pressurized water reactors (PWRs). These steels were irradiated with 1 MeV protons to doses between 1 and 10 dpa at 300 °C both with or without 15 appm helium pre-implanted at ∼100 °C. They were also irradiated at 340 °C, but only after 15 appm helium pre-implantation. Small heterogeneously distributed voids were observed in both alloys irradiated at 300 °C, but only after helium pre-implantation. The pre-implanted steels irradiated at 340 °C exhibited homogenous void formation, suggesting effects of both helium and irradiation temperature on void nucleation. Voids developed sooner in the SA304 alloy than CW316 alloy at 300 and 340 °C, consistent with the behavior observed at higher temperatures (>370 °C) for similar steels irradiated in the EBR-II fast reactor. The development of the Frank loop microstructure was similar in both alloys, and was only marginally affected by pre-implanted helium. Loop densities were insensitive to dose and irradiation temperature, and were decreased by helium; loop sizes increased with dose up to about 5.5 dpa and were not affected by the pre-implanted helium. Comparison with microstructures produced by neutron irradiation suggests that this method of helium pre-implantation and proton irradiation emulates neutron irradiation under PWR conditions.     

Swelling Behaviors in a Fuel Assembly for the Wrapping Wire and Duct Made of Modified 316 Austenitic Stainless Steel

Swelling behaviors in the wrapping wire and duct made of modified type 316 austenitic stainless steel were investigated in a fuel assembly irradiated in a fast breeder reactor. The temperature dependence of volumetric swelling was measured in the wrapping wire and the duct, and the peak temperatures of swelling were evaluated. The void distribution in the material was measured by microstructure observation with electron microscopy, and it was found that the voids prefentially grew near the surface. This phenomenon seemed to be caused by a surface effect on the neutron-irradiated materials.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

Dislocation loop evolution under ion irradiation in austenitic stainless steels

A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation.Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.     

Neutron irradiation-induced embrittlement in type 316 and other austenitic steels and alloys

The effects of neutron irradiation on the structures, high-temperature tensile and creep-rupture properties, deformation and fracture characteristics of austenitic alloys, particularly 316 type austenitic steels, are surveyed. The mechanisms proposed to explain the observed irradiation-induced changes in properties and behaviour are reviewed and the additional work required to resolve areas of uncertainty and difference is summarised.     

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.     

Effects of hold time and neutron irradiation on the low-cycle fatigue behaviour of type 316-CL and their consideration in a damage model

The creep fatigue behaviour of AISI type 316 L(N) plate material has been investigated in the temperature range of 450–750 °C by performing axial strain controlled tests with GRIM specimens. The creep and creep fatigue behaviour of austenitic stainless steel material is known to be prone to neutron irradiation-induced embrittlement. Therefore, the irradiation behaviour was studied by performing irradiation experiments in the High Flux Reactor (HFR) of Petten at 550 °C. A newly developed damage model for time-dependent damage was applied to describe the failure behaviour of AISI 316 L(N) in the cyclic tests performed.     

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.     

Structural,electrical and magnetic measurements on oxide layers grown on 316L exposed to liquid lead–bismuth eutectic

Fast reactors and spallation neutron sources may use lead–bismuth eutectic (LBE) as a coolant. Its physical, chemical, and irradiation properties make it a safe coolant compared to Na cooled designs. However, LBE is a corrosive medium for most steels and container materials. The present study was performed to evaluate the corrosion behavior of the austenitic steel 316L (in two different delivery states). Detailed atomic force microscopy, magnetic force microscopy, conductive atomic force microscopy, and scanning transmission electron microscopy analyses have been performed on the oxide layers to get a better understanding of the corrosion and oxidation mechanisms of austenitic and ferritic/martensitic stainless steel exposed to LBE. The oxide scale formed on the annealed 316L material consisted of multiple layers with different compositions, structures, and properties. The innermost oxide layer maintained the grain structure of what used to be the bulk steel material and shows two phases, while the outermost oxide layer possessed a columnar grain structure.     

Local mechanical properties of Alloy 82/182 dissimilar weld joint between SA508 Gr.1a and F316 SS at RT and 320 °C

The distributions of mechanical and microstructural properties were investigated for the dissimilar metal weld joints between SA508 Gr.1a ferritic steel and F316 austenitic stainless steel with Alloy 82/182 filler metal using small-size tensile specimens. The material properties varied significantly in different zones while those were relatively uniform within each material. In particular, significant gradient of the mechanical properties were observed near the both heat-affected zones (HAZs) of F316 SS and SA508 Gr.1a. Thus, the yield stress (YS) was under-matched with respect to the both HAZs, although, the YS of the weld metal was over-matched with respect to both base metals. The minimum ductility occurred in the HAZ of SA508 Gr.1a at both test temperatures. The plastic instability stress also varied considerably across the weld joints, with minimum values occurring in the SA508 Gr.1a base metal at RT and in the HAZ of F316 SS at 320 °C. The transmission electron micrographs showed that the strengthening in the HAZ of F316 SS was attributed to the strain hardening, induced by a strain mismatch between the weldment and the base metal, which was evidenced by high dislocation density in the HAZ of F316 SS.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

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.     

Hybrid PDMS-Silica-Zirconia materials prepared by γ-irradiation

Hybrids are a class of materials that enable the integration of organic and inorganic characteristics at the molecular level in a single material. In this way materials with unusual optical, mechanical or even bioactive properties are obtained, which are especially suitable for applications as sensors, non-linear optical (NLO) materials, lasers, selective membranes, catalysts and protective coatings. Sol-gel processing is often used for the preparation of hybrids, usually through the alkoxides method that leads to high purity products at relatively low temperatures. The authors have developed a new method for the preparation of these materials based on gamma irradiation of the precursors mixture. The precursors used are polydimethylsiloxane (PDMS), tetraethylortosilicate (TEOS) and zirconium propoxide (PrZr). The irradiations were performed using the ⁶⁰Co source at the Portuguese Gamma Irradiation Facility (UTR) located in the ITN campus at Sacavém. The materials at room temperature are macroscopically transparent, relatively flexible and amorphous. The details of the preparation of these hybrid materials by gamma irradiation are presented.     

The possible influence of order-disorder transformations on the swelling of 85Ni-15Cr nichrome alloy

The binary alloy 85Ni-15Cr swells during neutron irradiation in a manner that is quite unrepresentative of either simple austenitic alloys or pure nickel. The nickel-chromium system is known to exhibit ordering out of reactor and this alloy was used to test the concept that alloys which develop order exhibit lower swelling rates. When the neutron irradiation data are compared to that of ion irradiations conducted by Hudson and Ashby on Ni-Cr alloys, it is shown that chromium additions appear to depress swelling in nickel initially but also appear to suppress the tendency for swelling to saturate at high exposure. Below a transition temperature near 550°C the swelling is relatively sluggish and is quite insensitive to irradiation temperature. Above the transition temperature the swelling behavior is more complex but typical of austenitic alloys. The swelling transition temperature is thought to be related to the critical temperature for order-disorder transformation.