High Heat Flux Testing of B4C／Cu and SiC／Cu Functionally Graded Materials Simulated by Laser and Electron Beam

B4C，SiC and C,Cu functionally graded-materials(FGMs) have been developed by plasma spraying and hot pressing.Their high-heat flux properties have been investigated by high energy laser and electron beam for the simulation of plasma disruption process of the future fusion reactors.And a study on eroded products of B4C/Cu FGM under transient thermal load of electon beam was performed.In the experiment SEM and EDS analysis indicated that B4C and SiC were decomposed.Carbon was preferentially evaporated under high thermal load,and a part of Si and Cu were melted,in addition,the splash of melted metal and the particle emission of brittle destruction were also found.Fifferent erosive behaviors of carbon-based materials(CBMs)caused by laser and electron beam were also discussed.     

Angular properties of 6.13 MeV source photons penetrating lead,steel and concrete

Experimentally determined angular flux spectra for 6.13 MeV source photons from disc geometry penetrating shielding slabs of lead, steel and concrete are analysed. For all three shield materials, the proportion of photons scattered in a forward direction increases with increasing penetration thickness. At any given polar angle, the scattered photon properties decrease exponentially with increasing shield thickness. Angular exposure dose build-up factor is defined and angular contributions to the scalar build-up factor are presented. Changes in the distribution of photon energies relating to shield penetration thickness and polar angle are discussed. The influence of bremsstrahlung and positron annihilation photon sources are considered.     

无机吸附材料处理放射性废水中铀的研究进展

核能与核工业的发展不可避免地会产生大量的含铀放射性废水。随着材料科学的发展,新型吸附材料不断出现,吸附法处理放射性废水近年来得到了特别关注。目前研究的无机吸附材料可分为无机矿物、金属氧化物、碳基材料等。通过接枝、交联、等离子体、复合等技术处理,在吸附材料表面修饰含氧官能团或含氮官能团,可提高对铀的吸附性能和选择性。铀的吸附机理多表现为离子交换和表面络合作用。本文主要介绍了无机吸附材料处理放射性废水中铀的研究进展,并对今后的研究方向提出了建议。     

Sputtering of hydrocarbons by ion-induced breaking of chemical bonds

While the physical sputtering of atoms caused by keV and MeV ions has been studied extensively both by molecular dynamics (MD) simulation and experiments, the mechanisms leading to atom and molecule erosion at energies 1–100 eV are not very well understood. We now describe how low-energy hydrogen ions can cause erosion of carbon atoms and hydrocarbon molecules by entering the region of a carbon–carbon chemical bond and hence breaking it, a process we call ‘swift chemical sputtering'. In the particular case of hydrogen bombardment of carbon-based materials, we further show that this can lead to erosion yields far exceeding those expected for a physical sputtering process.     

Light emission from carbon-based materials under ITER relevant thermal shock loads

Light emission from carbon-based materials (fine grain graphite, CFC and silicon doped CFC) was observed during ITER relevant thermal shock loads by means of in situ optical diagnostics. The light emission which corresponds to particle release clearly indicated different particle release processes in the three materials. The differences were also found in the initiation temperatures of particle release and the surface morphology of the loaded areas. These results are related to the thermal stress in bulk materials. In addition to particle release, vapor cloud formation caused by thermal shock loads were observed as CII lines and lines from the C₂ Swan system. No Si lines but lines from SiC₂ molecules (Merrill-Sanford bands) were observed in Si doped CFC. This indicates that atomic silicon is not released under ITER relevant thermal shock loads.     

聚变堆实验包层模块第一壁温度场及热结构分析

铍相对于众多聚变反应堆的第一壁护甲材料,有着许多优点,这些优点使它和钨及碳基材料一起被选作国际热核聚变实验堆（ITER）第一壁的候选防护材料。对中国氦冷固态增殖剂实验包层模块（CHHCSBTBM）第一壁进行多场耦合模拟分析结果表明,使用表面热负荷模拟分析时,未考虑中子负载情况下,模拟分析结果与其它结果有较大出入,故使用表面热负荷模拟分析时必须考虑中子负载情况。而对第一壁热结构分析表明,铍保护板的应力超过了其许用应力,可以寻找其它铍合金或第一壁护甲材料以满足第一壁护甲材料热结构应力要求。     

Thermophysical studies of radwaste repository covers

This article is devoted to thermophysical studies of a multifunctional cover consisting of natural and artificial materials of a 2.8 m thick near-surface repository. The objectives were to determine the maximum depth of the frost zone and to evaluate the changes of the thermophysical properties of natural materials comprising the cover. The analysis showed that the depth of the frost zone does not drop below 0.6 m. The change of the thermophysical properties of the building materials used for a multifunctional cover is manifested as a decrease in the rate of advance of the temperature field. The thermal insulation properties of the repository are more than adequate – for total cover thickness 2.8 m the depth of the frost zone is 0.6 m.     

中国抗中子辐照钢的抗辐照设计与验证

聚变堆等未来先进核能系统要求材料在强流高能中子辐照下长期保持良好的结构稳定性和机械性能。为适应未来先进核能技术发展的需要,中国科学院核能安全技术研究所•凤麟团队牵头研发了具有我国自主知识产权的中国抗中子辐照钢--CLAM钢。CLAM钢的设计考虑了未来核能清洁性的要求,以及苛刻服役环境中材料抗辐照、耐高温、耐腐蚀等性能要求。通过中子学计算分析设计了低活化成分范围,基于选择性纳米相析出进行了抗辐照、耐高温性能优化设计。针对材料的抗辐照性能,利用国内外中子、离子、电子及等离子体辐照设施开展了系列辐照考验研究,通过多角度表征辐照前后材料的微观结构和宏观性能,综合评估了材料的辐照性能,并与国际上同类材料在相近或相同条件下的辐照性能进行了对比分析,结果表明CLAM钢具有良好的抗辐照性能。     

P92钢高温低周疲劳的实验研究

由于高的热效率和简单的系统组成,超临界水堆(SCWR)被认为是第四代核反应堆的一种选择。超临界水堆的关键问题之一是核心部件尤其是燃料组件包壳的材料。这些材料在高温下的力学性能、腐蚀和应力腐蚀开裂敏感性以及抗辐射性能等对核电厂的安全运行至关重要。本文对SCWR包壳候选材料的F/M类材料P92钢进行了高温低周疲劳实验研究。实验温度为600和650℃,控制方式为总应变控制,应变范围均为±0.2%～±0.6%。实验结果表明,在两种温度下,P92钢均为循环软化材料,但未出现循环稳定现象。由于温度升高,塑性增强,P92钢在650℃下的宏观裂纹出现周次比率随应变范围的增加,下降比较平缓,且650℃下的失效寿命显著高于600℃下的失效寿命。并得到了两种温度下的稳定循环应力-塑性应变的关系以及循环失效寿命和应变的关系。     

Modeling of chemical erosion of graphite due to hydrogen by inclusion of chemical reactions in SDTrimSP

Erosion of carbon-based materials plays a very crucial role in determining its candidature as a potential plasma facing material in fusion devices. Carbon erosion yield shows strong dependence on energy, flux of the incoming ions and the target temperature. Various experiments and theoretical models have made successful attempts to understand some aspects of the erosion mechanism of carbon bombarded by hydrogen ions. They are valid for a particular parameter range. None of the models alone can explain the temperature, energy and flux dependence of erosion simultaneously. In the present work different aspects of erosion mechanism are parameterized. A new model which takes into account the basic merits of different theoretical models and the various parameterized quantities is embedded in the SDTrimSP [1], [2] program. This upgraded SDTrimSP enables us to calculate the physical and chemical sputtering in a single model for various parameter ranges. The results obtained by the upgraded SDTrimSP are in good agreement with experimental data.     

Magnetic order in proton irradiated graphite: Curie temperatures and magnetoresistance effect

Defect induced magnetic order is a new phenomenon in material science that refers to the triggering and manipulation of magnetic order and magnetic moments in nominally non-magnetic materials by lattice defects and/or non-magnetic add atoms. A noticeable example of this effect is the magnetic order at room temperature produced by proton irradiation of graphite. In this work we have managed to increase the ferromagnetic signal by cooling the graphite samples down to 110 K during proton irradiation, diminishing in this way annealing effects. SQUID measurements of the magnetization show a fluence dependent Curie temperature. The longitudinal magnetoresistance shows an irreversible behavior similar to that found in ferromagnetic films indicating spin/domain reorientation effects. The observed magnetoresistance effects and Curie temperatures above room temperature are promising facts that may lead to useful carbon-based devices in the near future.     

Development of new nuclear shielding materials containing vitrified colemanite and impregnated polymer

New shielding materials were developed by using vitrified colemanite in powder form in cement phase and as particles in coarse aggregate in concrete. Vitrification was done to decrease the deleterious effects of colemanite on the setting of cement and thus to increase the boron content in the shielding wall to improve thermal neutron absorption. The boron oxide content of concrete was shown to increase to 15 wt% which is about 5–6 times more than the boron oxide content of conventional nuclear reactor concrete walls. Iron chunks were partly used in coarse aggregate to compensate for the decrease in density due to vitrified colemanite, and thus the improve attenuation of gamma rays. Polymer impregnation of hardened concrete improves its mechanical and shielding properties. By using a computer shielding program it was found that the materials developed usually have better shielding properties than the conventional shielding walls.     

聚变堆金属材料中子辐照多尺度计算模拟

中子辐照条件下材料结构与性能是中国聚变工程实验堆(CFETR)以及未来聚变反应堆工程设计的重要依据.钨材料是CFETR拟全面使用的壁材料,但中子辐照导致钨硬度升高和韧性大幅下降,严重影响材料的服役性能,进而影响CFETR运行的安全性和稳定性.在目前缺乏聚变中子源进行辐照实验的情况下,开展聚变堆材料中子辐照模拟研究显得愈...     

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon-carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon-carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10¹² n/cm²/s at temperature of 40 °C. The fluence was 2.52 × 10¹⁶ n/cm². These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

不同材料中子反射与屏蔽效应研究

利用Monte Carlo粒子输运计算程序Super MC对厚度1-5 cm的多种材料进行中子反射和屏蔽性能分析计算。这些材料包括金属材料铍、铅、铜、含硼钢以及~(238)U和非金属材料聚乙烯、氢化锂、混凝土,中子能段选取10~(-5) e V-20 MeV。结果显示,中子反射能力和屏蔽性能都会随着材料厚度而增加,但增加的幅度逐渐减小。铍和聚乙烯在中子反射和屏蔽方面性能优越,而常用来屏蔽γ射线的铅在这两方面性能都是8种材料中最差的。~(238)U只在材料厚度很小时性能卓著,随着材料厚度增加,其性能便远不如大部分材料。考虑到聚乙烯的力学性能较差,在屏蔽材料的选择上有很大的限制,所以在8种材料中,铍的综合性能相对较好。     

Thermophysical properties of MgO, UO2, their eutectic solution and slurry of liquid-solid mixtures, concrete, sodium, stainless steel and debris beds for use in molten pool penetration of MgO substrate

We have prescribed various thermophysical and transport properties to describe various thermal states of the materials of interest such as MgO, UO₂, stainless steel, sodium, and concrete undergo during post accident heat removal (PAHR) in an ex-vessel cavity lined with MgO bricks. A number of properties, especially of molten MgO, had no experimental determination and therefore, by necessity, these were prescribed through available “best” estimates. We have also included the equivalent properties of various “composite” materials such as debris beds with a prescribed composition, solutions, and slurries to describe their participation in various thermophysical phenomena of interest in PAHR.     

低能电子辐照对环氧树脂性能的影响

为了研究低能电子辐照对环氧树脂的体积电阻率、邵氏硬度、拉伸强度和官能团结构的影响,本文在电子辐照能量为30 keV,注量率1×10¹¹ cm^-2•s^-1,总注量为1.6×10¹⁴ cm^-2,真空度10^-6 Pa条件下,结合国家标准对辐照前、后环氧树脂材料的机械性能和结构进行表征。结果表明,辐照后环氧树脂材料的体积电阻率、邵氏硬度、拉伸强度等宏观物理性能均有下降。傅里叶红外光谱图显示环氧树脂主要官能团强度降低,产生的•H、•OH等自由基与聚合物分子上的羟基与氢结合。研究结果对环氧树脂材料在辐射环境中的使用具有重要意义。     

应用低温等离子体破碎岩石

一般地,破岩方法可分为机械能破岩和热能破岩两种,随着科学技术的发展,热能破岩的方法和技术已在很多领域应用,特别是利用火钻焰流开采石材在很多矿山广泛得到了应用.从实际应用出发,重点介绍了低温等离子体破岩、凿岩的原理和方法,等离子装置的结构及其特点.     

Toughness behaviour of the steel X 20 CrMoV 12 1 — Results of optimizing the base- and weld metal

The 12% Cr steels are frequently used in German power plants for tubings, pipes, rotors, and blades. The maximum operating temperature is limited by their creep strength properties to about 550°C. There are applications at even higher temperatures. Sufficient materials toughness is required for the base metal and weld metal to withstand sudden load changes. This is of special interest for use in nuclear power plants. Under operating conditions at elevated temperatures microstructural changes occur which greatly influence the toughness properties of both base metal and weld metal. This paper presents the results of ageing treatments at 550°C, carried out with a 12% Cr steel (DIN X 20 CrMoV 12 1) specifically optimized for toughness. The decrease in toughness is already evident at ageing times as low as 1000 h for conventional and optimized material. This drop in toughness is tentatively explained by differences in grain sizes and carbide content (M₂₃C₆ carbides). Detailed investigations indicate that additional carbide precipitation may significantly contribute to the decrease in toughness.