在役反应堆压力容器退火处理技术的应用

介绍了核电站在役反应堆压力容器退火处理技术的原理及其主要影响因素,讨论了退火后材料的韧性恢复程度与再辐照脆化速度两项退火处理的关键参数,展示了退火处理的工程实施方案,比较了俄罗斯与美国设计的反应堆压力容器在实施退火处理时的主要差异项,最后提出了我国现阶段在条件成熟时可适度开展一些必要基础研究工作的建议。     

International Nuclear Safety Center Database on Thermophysical Properties of Reactor Materials

The International Nuclear Safety Center (INSC) database has been established at Argonne National Laboratory to provide easily accessible data and information necessary to perform nuclear safety analyses and to promote international collaboration through the exchange of nuclear safety information. The INSC database, located on the World Wide Web at http://www.insc.anl.gov, contains critically assessed recommendations for reactor-material properties for normal operating conditions, transients, and severe accidents. The initial focus of the database is on thermodynamic and transport properties of materials for water reactors. Materials that are being included in the database are fuel, absorbers, cladding, structural materials, coolants, and liquid mixtures of combinations of UO₂, ZrO₂, Zr, stainless steel, absorber materials, and concrete. For each property, the database includes (1) a summary of recommended equations with uncertainties; (2) a detailed data assessment giving the basis for the recommendations, comparisons with experimental data and previous recommendations, and uncertainties; (3) graphs showing recommendations, uncertainties, and comparisons with data and other equations; and (4) property values tabulated as a function of temperature.     

Selection of Materials for Pressure Vessels and Chemical Plants

The selection of materials for pressure vessels and chemical plants depends on a number of factors such as operating pressure, operating temperature, operating medium, regulations in force in the country of the plant user concerned and manufacturing possibilities. The essay clearly explains how the above specified factors individually influence the selection of materials. The article also deals with the ranges of application of certain material groups such as unalloyed and low-alloy steels, fine-grained steels, austenitic chromium-nickel steels, unalloyed ferritic chromium steels and other materials. The article closes with remarks on the operational safety of pressure vessels.     

Structure and Corrosion-Mechanical Properties of the Surface Layers of Steels after Laser Alloying

We perform the thermodynamic analysis of the solubility of alloying elements and stability of possible secondary phases of chromium and chromium-nickel steels in the components of the Li₁₇Pb₈₃ eutectic heat carrier and reveal the negative effect of the lead melt on the long-term strength and plasticity of the major part of these steels. It is proposed to improve the stability of the characteristics of structural steels promising for nuclear and thermonuclear power engineering by the laser alloying (modification) of the surfaces with Zr, Nb, or Si and B. By analyzing the changes in the microhardness, structure, and thermoelectromotive force of the corresponding coatings on holding in the eutectic melt at 350 or 500°C for 10⁴ h, it is shown that the procedure of laser alloying may result in the formation of layers with long-term protective action.     

Assessment of In-Service Degradation of Mechanical Properties of NPP Reactor and Piping Materials

An experimental investigation has been carried out to clarify the influence of operating factors on static and cyclic strength of NPP reactor vessels and piping. The authors present the results of analysis of crack growth resistance characteristics of the base material and welded joints in WWER-1000 reactors (steel 15Kh2NMFA) in fatigue failure. The mechanisms of thermal aging of steels of NPP piping in long-term operation (up to 10,000 hours) have been studied.     

聚变堆中面向等离子体材料的研究进展

受控热核聚变能是公认的可以有效解决人类未来能源需求的主要途径之一,经过多年的努力,其研究已经取得很大进展,进入了从物理可行性向工程可行性的验证阶段.决定核聚变能未来发展的一个关键问题是相关的材料问题,尤其是面向等离子体材料的发展.评述了国内外目前核聚变实验装置中面向等离子体材料的研究进展.     

反应堆压力容器用SA508Gr.4N钢的热变形行为

利用Gleeble-1500D热模拟试验机,在温度为1050~1250℃、应变速率为0.001~0.1s-1、真应变量0.16的条件下,研究和分析SA508Gr.4N钢高温塑性变形及动态再结晶行为。结果表明:SA508Gr.4N钢的高温真应力-应变曲线主要以动态再结晶为特征,峰值应力随变形温度的降低或应变速率的升高而增加,属于温度和应变速率敏感材料;在真应力-应变曲线的基础上,建立材料热变形本构方程,较好地表征了材料高温流变特征,其热激活能为383.862kJ/mol;其硬化率-应力(θ-σ)曲线均呈现拐点且-dθ/dσ-σ曲线出现极小值;临界应变随应变速率的增大与变形温度的降低而增加,且临界应变(εc)与峰值应变(εp)之间具有一定相关性,即εc/εp=0.517;临界应变与Z参数之间的函数关系为εc=8.57×10-4 Z0.148。     

轻水反应堆（LWR）用包壳材料研究进展

轻水反应堆（LWR）是国际上多数核电站采用的堆型。锆具有良好的加工性能,优良的机械性能,较高的熔点、优异的耐蚀性能及核性能,被用作燃料包壳和堆芯结构材料,是发展核电及核动力舰船不可替代的关键结构材料和功能材料。随着核电技术的发展,对堆芯包壳材料性能提出了更高的要求,综述了核用锆合金包壳材料的国内外研究和使用现状以及新型SiC包壳材料的研发现状。总体来说,锆合金在未来几十年内仍是核反应堆包壳材料的主要用材,开展新合金的研发,不断提升锆合金的性能是世界各国研究者共同的目标;适时加大投入力度,强化条件建设,就能加快具有国内自主知识产权锆合金的产业化步伐,可最终实现核电及核动力用锆合金材料的自主化;SiC材料具有更高的熔点、更好的耐腐蚀性能,是一种极具应用潜力的材料,有可能成为第4代核反应堆的包壳材料,但还需投入大量研究。     

Reactor Graphites Based on Inexpensive Raw Materials

Data are presented on the performance characteristics and properties of Russian-produced graphites molded with inexpensive fillers—electrode, schistose, and pitch cokes. Replacement of petroleum coke, used earlier, by these cokes ensures performance parameters that meet the relevant specifications.     

NIST Materials Properties Databases for Advanced Ceramics

The NIST Ceramics Division maintains two databases on the physical, mechanical, thermal, and other properties of high temperature superconductors and structural ceramics. Crystallographic data are featured prominently among the physical property data and serve several important functions in the classification and evaluation of the property values. The scope of materials, properties, and data evaluation protocols are discussed for the two databases.     

浮空器用柔性复合材料的力学性能

通过试验研究了某浮空器蒙皮用柔性复合材料的基本力学性能和断裂力学性能。通过单轴拉伸试验研究了该材料经向和纬向试样的力学特性,结果发现两者的弹性模量均在10GPa左右,复合材料的柔软度很高;而经向拉伸试样的抗拉强度比纬向拉伸试样的高出约32％;通过改变切口相对宽度的撕裂试验发现撕裂试验的载荷一位移曲线和裂纹扩展特性基本表现为脆性断裂特性,试验结果和根据弹塑性断裂理论计算得到的结果相一致,即随着切口相对宽度的增加,柔性复合材料的极限载荷降低。     

聚变堆装置中面向等离子体材料钨涂层的研究进展

钨具有高的熔点、不与氚发生共沉积、与等离子体好的兼容性和低的腐蚀率等优点,是最有前景的一种面向等离子体材料.为了解决面向等离子体材料的制备及其与热沉材料连接问题,涂层技术在实验聚变堆装置中得到广泛应用.评述了目前实验聚变堆装置中面向等离子体材料钨涂层的研究进展.     

Effect of Inhibitors on the Corrosion-Mechanical Fracture of Steels in Hydrogen-Sulfide Media

We have studied the effect of Naftokhim-3, Naftogaz-1, Kaspii-2, Tyumen-2, and Travis TKh-1102 inhibitors on the corrosion-mechanical strength of pipe steel in hydrocarbon and aqueous media containing hydrogen sulfide. We have established that these inhibitors increase the plasticity and fracture energy of steel and, at the same time, worsen its strength properties. In addition, they decrease the crack growth rate under cyclic loading. Naftokhim-3 and Naftogaz-1 inhibitors have the highest efficiency in aqueous hydrogen-sulfide media and Naftokhim-3 in hydrocarbon media.