钇离子注入不锈钢的抗腐蚀研究

叙述了在复旦大学加速器实验室３０ｋｅＶ同位素分离器上建立的一套离子注入装置。在该装置上已获得几十微安的Ｙ、Ｃｅ、Ｈｏ、Ｓｍ、Ｎｄ、Ｌａ、Ｙｂ等稀土离子束流。被离子注入的面积为φ２０ｍｍ，表面均匀性好于３％，并对Ｙ离子注入不锈钢以改善其抗腐蚀性能进行了初步研究。     

钼离子注入硅薄层硅化钼的合成

用大束流密度的钼金属离子注入硅,能够直接合成性能良好的薄层硅化物。随着束流密度的增加,硅化钼生长,薄层硅化物的方块电阻Ｒｓ明显下降,当束流密度为０．５Ａ／ｍ＾２时,Ｒｓ达到上值９０Ω,说明连续的硅化物已经形成。Ｘ衍射分析结果表明,注入层中形成了３种硅化钼Ｍｏ３Ｓｉ、Ｍｏ５Ｓｉ３和ＭｏＳｉ２。经过９００℃退火后,Ｒｓ下降至４Ω,电阻率可小到０．１６μΩ．ｍ,说明硅化钼薄层质量得到了进一步的改善。大束     

辐照导致奥氏体不锈钢内位错环演化行为模拟

为了描述和预测奥氏体不锈钢材料在辐照条件下位错环的演化行为,本文基于平均场速率理论结合分子动力学方法建立辐照诱导奥氏体不锈钢内位错环演化的物理模型,模拟电子、中子辐照诱导奥氏体不锈钢内位错环的演化行为,模拟结果与实验结果吻合很好。在此基础上探究了中子辐照产生的存活缺陷对材料内位错环演化行为的影响,研究表明级联碰撞过程中的缺陷存活率、缺陷成团率以及四间隙原子团簇所占的团簇份额是影响位错环演化行为的主要参数,而双间隙原子团簇份额和三间隙原子团簇份额之间比例对位错环的演化没有影响。     

反应堆奥氏体不锈钢锻件的超声试验

一、前言奥氏体不锈钢除有优良的核性能外,还有良好的耐腐蚀性和冷热加工性,在常温和低温下,有优良的塑性和韧性,没有磁性,因此,核反应堆中一些零部件,由奥氏体不锈钢制成,生产过程中产生各种缺陷,如原材料中的缩孔、疏松、夹渣、锻造裂纹、白点 折迭和热处理裂纹等,这些缺陷的存在,降低核动力装置部件的机械性能,不消除隐患,易造成事故。为了使反应堆安全、可靠运行,对奥氏体不锈钢锻件需进100%的超声无损检测。     

核电站一回路奥氏体不锈钢铸件热老化的微观力学性能变化

在核电站一回路中,奥氏体不锈钢铸件(CASS)在运行温度下长期工作除了受到辐照损伤外.还将面临热老化问题.本文对不同时间热老化后的CASS样品进行微观力学性能分析,并结合微观组织中铁素体的含量及分布对CASS的性能变化进行了分析,结果表明微观力学性能的变化和常规的宏观力学性能变化的规律不尽相同.丰富了CASS经老化后的力学性能数据,为CASS的安全运行提供参考.     

奥氏体不锈钢的显微组织及氦脆断口观察

利用电子显微镜对几种奥氏体不锈钢的形变和时效后的显微组织作了研究,在低应变下加工硬化主要依靠位错密度的增加,高应变下还出现大量形变诱发组织。时效引起碳化物的析出,基体还有再结晶的趋势。此外,还对经α粒子辐照后的氦脆断口作了观察。     

超级奥氏体不锈钢UNS08367的焊接性能

近年来,超级奥氏体不锈钢凭借其优异的耐C L-腐蚀性能被广泛应用。文章通过对UNS08367全焊缝金属的拉伸试验以及对焊缝及热影响区的冲击试验,分析焊接对UNS08367力学性能的影响;通过对U NS08367焊缝及热影响区焊态以及固熔处理态耐点蚀和耐缝隙腐蚀试验,分析焊接对UNS08367耐腐蚀性能的影响。     

He离子注入不锈钢中的温度影响

用增强质子背散射、SEM、TEM和显微硬度测量等手段,研究了He离子注入316L型不锈钢中的行为,离子束能量和剂量区域分别为30—170keV和10~(14)-3×10~(17)He~ /cm~2,靶温和后退火的温度区域为77—673K。研究了温度对He捕获、He泡结构、He泡的生长和合并等特性的影响。测量了注氦层显微硬度的变化。     

酸性硫酸根离子介质中奥氏体不锈钢的应力腐蚀行为研究

用慢应变速率法(SSRT)研究奥氏体不锈钢316Ti和316NG在酸性硫酸根离子介质中的应力腐蚀(SCC)行为.结果表明在酸性硫酸根离子环境中,316Ti产生应力腐蚀的临界浓度比316NG的高,但316NG的裂纹扩展速率比316Ti大.用电化学阳极极化法比较了两种奥氏体不锈钢在酸性硫酸根离子介质中的抗腐蚀性能的差异,结合微观分析,探讨了SO2-4离子对奥氏体不锈钢的SSRT-SCC损伤机理.     

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.     

Time-temperature-sensitization diagrams and critical cooling rates of different nitrogen containing austenitic stainless steels

Nitrogen-alloyed 316L stainless steel is being used as structural material for high temperature fast breeder reactor components with a design life of 40 years. With a view to increase the design life to 60 years and beyond, high nitrogen stainless steels are being considered for certain critical components which may be used at high temperatures. Since carbon and nitrogen have major influence on the sensitization kinetics, investigations were carried out to establish the sensitization behaviour of four heats of 316L SS containing (i) 0.07%N and 0.035%C, (ii) 0.120%N and 0.030%C, (iii) 0.150%N and 0.025%C and (iv) 0.22%N and 0.035%C. These stainless steels were subjected to heat treatments in the temperature range of 823-1023 K for various durations ranging from 1 h to 500 h. Using ASTM standard A262 Practice A and E tests, time-temperature-sensitization diagrams were constructed and from these diagrams, critical cooling rate above which there is no risk of sensitization was calculated. The data established in this work can be used to select optimum heat treatment parameters during heat treatments of fabricated components for fast reactors.     

应变幅对国产锻造奥氏体不锈钢环境疲劳寿命影响的试验研究

为验证模拟压水堆核电站冷却剂服役环境对国产锻造主管道用奥氏体不锈钢疲劳寿命的影响,采用高温高压循环水疲劳测试系统对从产品锻件取样加工后的标准试样进行了低周疲劳试验,分析了试验数据与美国机械工程师学会(American Society of Mechanical Engineers,ASME)规范平均/设计疲劳曲线的关系,获得了应变幅对奥氏体不锈钢环境疲劳寿命的影响规律,并初步评价了ASME规范设计疲劳曲线和环境疲劳修正系数的适合性。     

聚苯胺的离子注入掺杂研究

研究了聚苯胺的离子注入掺杂对导电性的影响。用红外和紫外谱法探讨了聚苯胺的氧化态，本征态和还原态经离子注入后，其结构可能发生的变化。结果表明，聚苯胺的三种态的离子注入均为还原过程。讨论了离子注入掺杂的导电机制。     

背散射分析钇离子注入H13钢的抗氧化特性

用背散射技术测量了不同注入剂量和大束流密度的Y~+注入H_(13)钢的浓度分布,并研究了注入样品的抗氧化特性。结果表明,Y的浓度分布和峰值浓度的变化与注入剂量和束流密度密切相关。Y峰值浓度大小对抗氧化特性有直接的影响。Y~+注入钢改变了钢的氧化模式。氧的扩散受到控制。高温(800℃)氧化10min即可形成抗氧化的氧化皮。Y~+的注入可使氧化速率下降8倍。     

离子束混合及离子注入陶瓷材料表面改性研究概述

对离子注入陶瓷材料引起的辐照损伤和材料力学性能、摩擦学性能的改善及陶瓷基体上金属薄膜的离子束混合增强粘着研究的进展进行了综述。     

离子注入α－Al_2O_3陶瓷材料的表面改性研究

研究了α－Ａｌ２Ｏ３单晶注入Ｎ＋后的力学性能变化。结果表明，注入剂量为１×１０１７Ｎ＋／ｃｍ２时，Ａｌ２Ｏ３的显微硬度提高了９２％；注入剂量为３×１０１７Ｎ＋／ｃｍ２时，在Ａｌ２Ｏ３表面形成非晶层，导致表面软化，显微硬度只是基体的５０％左右．离子注入在样品表面产生了高达￣１１５０ＭＰａ的压应力，材料的断裂韧性的改善与此有关，实验亦发现Ｎ＋离子注入后Ａｌ２Ｏ３的断裂韧性提高了９５％左右．ＳＥＭ分析同样证明Ｎ＋注入ａ－Ａｌ２Ｏ３后，其表面力学性能确实有所改善。     

Simulation of damage induced by ion implantation in Lithium Niobate

A simulation tool has been developed to engineer the damage formation in Lithium Niobate by ion irradiation with any atomic number and energy. Both nuclear and electronic processes were considered and, in particular, the dependence on the ion velocity of the electronic excitation damage efficiency has been taken into account. By using this tool it is possible both to draw damage nomograms, useful to qualitatively foresee the result of a given process, and to perform reliable simulations of the defect depth profiles, as demonstrated by the good agreement with the experimental data available in the literature.