高温He离子辐照GH3535合金的损伤效应

采用500 keV的He离子在750 ℃下对GH3535合金样品进行辐照,然后利用掠入射X射线衍射（GIXRD）、透射电子显微镜（TEM）和纳米压痕仪分别对样品的氦泡和位错环辐照缺陷的演化及纳米硬度的变化进行了研究。结果表明,GH3535合金晶格辐照后发生了轻微畸变;离子辐照在样品中形成了大量尺寸为2～5 nm的氦泡和位错环。辐照产生的氦泡和位错环等缺陷在基体中钉扎位错,从而使材料产生了辐照硬化现象,样品硬度随辐照剂量的增加而增大。当辐照剂量达2×1016 cm-2时,辐照样品发生了明显的硬化饱和现象,利用Nix Gao模型计算得此时的硬化程度为64%。     

FeCr合金中Cr含量对微观结构影响的原子尺度模拟研究

低温辐照脆化是影响铁素体/马氏体（F/M）钢服役的主要问题之一。F/M钢低温辐照脆化的主要机理是辐照产生的纳米缺陷（如位错环、α′相(富Cr团簇)等）阻碍位错运动。本文利用分子动力学方法和迈氏蒙特卡罗方法对F/M钢模型材料--FeCr合金（Fe7%Cr、Fe9%Cr、Fe14%Cr）中Cr元素析出成团簇及在位错环上偏析的机理进行研究,并分析Cr团簇析出与合金成分的关系以及位错环尺寸、位错环类型和合金中Cr含量对位错环上Cr偏析量的影响。模拟结果表明：热力学模拟后,高Cr含量（>9%)的FeCr合金中会析出Cr团簇,且基体内Cr含量越高,析出的Cr团簇尺寸越大;在所研究的3种FeCr合金中,受位错环张应力场作用,合金元素Cr均会在位错环的外围偏析,且FeCr合金中Cr含量越高,Cr在位错环上偏析量越高。低Cr的FeCr合金中Cr对其辐照硬化的影响需考虑位错环上Cr偏析的影响,高Cr的FeCr合金中Cr元素对其辐照硬化的影响需综合考虑Cr团簇及位错环上Cr偏析。     

2MeV质子辐照对Zr-4合金显微组织的影响

通过密西根大学离子束表面改性和分析实验室的大束流加速器研究了Zr-4合金的质子辐照效应。结果表明:当原子离位损伤率约1×10-5 dpa/s,在350癈 质子辐照损伤分别达到2dpa、5dpa和7dpa时,辐照后位错环的密度分别为7×1021/m3、8×1021/m3、15×1021/m3,尺寸分别为7nm、11nm和11nm,表明位错环的密度和尺寸随质子辐照注量有增加的趋势。辐照前后的明场像、高分辨相和电子衍射花样均表明,在350癈 2MeV的质子辐照没有使锆4合金中的hcp-Zr(Cr,Fe)2和fcc-ZrFe2沉淀相发生非晶化转变。     

C276合金质子辐照损伤模拟及活化分析

C276合金是先进核电站燃料元件包壳的候选材料之一。本工作采用TRIM程序分别计算10和20MeV质子辐照C276所产生的辐照损伤,比较分析能损、离位原子、DPA等参数分布。同时使用FISPACT-2007程序进行活化计算,对放射性活度、衰变余热及接触剂量率等参数进行了详细分析。结果表明：辐照损伤主要来自电子能损的贡献,高能质子与靶原子发生碰撞的几率较低。C276经同种能量质子辐照后,活化特性随着辐照时间的增长而增加。辐照时间相同时,高能质子会对材料产生更大的影响。本工作为后续的辐照损伤分子动力学模拟及计划开展的质子辐照实验提供支持。     

锆合金原位离子辐照条件下的位错环演化研究

位错环演化是核用锆合金辐照组织演化的主要特征之一,对合金辐照后的力学性能（强度、塑性等）有着决定性的影响。目前,锆合金辐照位错环演化的实验研究主要基于离位中子或离子辐照,无法直接观察位错环的演化过程。为了更深入地理解锆合金辐照下的微观组织演化,本工作采用先进的原位离子辐照实验方法,实时观察Zr-2合金位错环的演化过程,揭示不同辐照损伤剂量和温度对演化过程的影响规律,并结合弥散障碍物硬化模型对合金的辐照硬化性能进行了评估,验证了原位离子辐照用于研究锆合金包壳材料辐照后位错环演化和力学性能评价的可行性和先进性。      

位错环上的Cr偏析对辐照硬化影响的原子尺度模拟研究

低温辐照脆化是影响低活化铁素体/马氏体（RAFM）钢服役的主要问题之一。RAFM钢低温辐照脆化的主要机理是辐照产生的纳米缺陷（如位错环、析出物等）阻碍位错运动。本文利用分子动力学方法研究了bcc-Fe内刃型位错线与1/2〈111〉间隙位错环的相互作用,并对比分析了Cr偏析在位错环上对其硬化的影响。研究结果表明：刃型位错线挣脱位错环所需临界剪切应力（CRSS）与位错环的伯格斯矢量有关;在本文所研究条件下,在一定温度范围内,Cr偏析在位错环上会使得位错线挣脱所需CRSS增加,引起硬化增强。     

氦离子辐照烧结碳化硅损伤效应研究

利用中国科学院近代物理研究所320 kV高压平台提供的氦离子辐照烧结碳化硅,辐照温度从室温到1 000 ℃,辐照注量为1015~1017 cm-2。辐照完成后,进行退火处理,然后开展透射电子显微镜、拉曼光谱、纳米硬度和热导率测试。研究发现,烧结碳化硅中氦泡形核阈值注量低于单晶碳化硅。同时,氦泡形貌和尺寸与辐照温度、退火温度有关。另外,对辐照产生的晶格缺陷、元素偏析进行了研究。结果表明,辐照产生了大量的缺陷团簇,同时氦泡生长也会发射间隙子,在氦泡周围形成间隙型位错环。在晶界处,容易发生碳原子聚集。辐照导致材料先发生硬化而后发生软化,且热导率降低。     

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。     

高温He离子辐照GH3535合金的损伤效应

采用500 keV的He离子在750℃下对GH3535合金样品进行辐照,然后利用掠入射X射线衍射(GIXRD)、透射电子显微镜(TEM)和纳米压痕仪分别对样品的氦泡和位错环辐照缺陷的演化及纳米硬度的变化进行了研究。结果表明,GH3535合金晶格辐照后发生了轻微畸变;离子辐照在样品中形成了大量尺寸为2～5 nm的氦泡和位错环。辐照产生的氦泡和位错环等缺陷在基体中钉扎位错,从而使材料产生了辐照硬化现象,样品硬度随辐照剂量的增加而增大。当辐照剂量达2×10~(16) cm~(-2)时,辐照样品发生了明显的硬化饱和现象,利用Nix-Gao模型计算得此时的硬化程度为64%。     

Zr-Sn-Nb合金质子辐照效应研究

摘要:通过HZB串列加速器用18MeV质子研究了ZrSnNb合金的辐照效应。结果表明,经1.53×1014cm-2注量的质子辐照后,辐照产生的缺陷使样品的显微硬度升高导致辐照硬化;断口分析发现辐照后在样品的一端韧窝明显变小出现脆化现象,被认为是由于质子辐照离子化损伤产生在其射程的末端——与TRIM96程序计算的结果一致。     

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.     

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.     

Mechanical properties and microstructure in neutron-irradiated nickel-based alloys and stainless steels for supercritical water-cooled-reactor fuel cladding

ABSTRACT

To investigate the irradiation behavior of mechanical properties and microstructural changes of commercial Ni-based alloys and improved stainless steels, a neutron-irradiation experiment was performed at the Joyo reactor, and post-irradiation examinations with tensile tests and TEM observations were carried out. The room-temperature tensile tests showed that all specimens that were irradiated at 485°C exhibited significant hardening and ductile behavior, especially in alloy 625. The irradiation hardening of all specimens irradiated at 668°C was less than that of specimens irradiated at 485°C. The fine-grained stainless steel, T3 and the Zr-added stainless steels, H1 and H2 showed good mechanical-property performance with keeping ductility after neutron irradiation. Most alloys and steels showed ductile behavior on the fracture surface except for alloy 625 specimen. The TEM observations showed that a high density of tangled dislocations and irradiation-induced defect clusters formed in the stainless steels and Ni-based alloys irradiated at 485°C. At 668°C, the material microstructures coarsened and their dislocation density decreased significantly. Long rod-like precipitates of Zr(Cr, Fe) compounds formed in the H1 and H2 steels that were modified with Zr. The yield stress drop of T3 steel in tensile stress was observed and is caused by grain-size coarsening at an irradiation of 668°C.     

Effects of an alloying element on a c-component loop formation and precipitate resolution in Zr alloys during ion irradiation

It is important to clarify the mechanisms of the dislocation loop formation, dissolution of precipitates to understand the degradation behavior of the fuel cladding tubes in light water reactors (LWR) under neutron irradiation. In this study, 3.2 MeV Ni ion irradiation was carried out at 400°C on Zircaloy-2 and two types of model alloys with and without Fe (Zr-1.5Sn-0.3Fe and Zr-1.5Sn). To understand the effects of hydrogen, 60 and 300 ppm pre-injected Zircaloy-2 samples were also irradiated. The microstructure was observed with a conventional transmission electron microscopy. Additionally, the dissolution of precipitates and the enrichment of the alloying element due to irradiation were analyzed using a spherical aberration (Cs)-corrected high-resolution analytical electron microscope. After ion irradiation at 400°C, the dissolution of Fe-enriched precipitates and the c-component dislocation loops were observed in the region of peak ion damage. Observations by STEM-EDS showed that Fe atoms were enriched in the c-component dislocation loops. On the contrary, the c-component dislocation loops were detected in Fe-containing alloys (Zircaloy-2 and Zr-1.5Sn-0.3Fe alloy) but were not in the Zr-1.5Sn alloy. These results indicate that the dissolution of Fe-enriched precipitates and the enhanced formation of c-component dislocation loops are essential for the degradation of LWR fuel cladding under irradiation.     

Neutron irradiation hardening and microstructure changes in Fe-Mn binary alloys

Irradiation hardening and microstructure changes in Fe-Mn binary alloys were investigated after neutron irradiation at 290 °C and up to 0.13 dpa. Significant irradiation hardening comparable to that of Fe-1 at.%Cu alloy was observed in Fe-1 at.%Mn alloy. Manganese increases the number density of dislocation loops, which contributed to the observed irradiation hardening. Manganese serves as a nucleus of the loop by trapping interstitial atoms and clusters, preventing 1D motion of the loops.