高放废物地质处置中的工程材料

凡人类从事于与核材料有关的许多生产、生活活动均可能产生不同活度的放射性废物.高放废物由于具有放射性水平高、发热量大、核素寿命长等特点,其安全处置倍受全球科学家和广大公众所重视.目前深地质处置被国际上公认为处置高放废物的最有效可行的方法.借鉴已有研究成果,我国采用多重工程屏障系统(包括废物固化体、废物罐及其外包装和缓冲/回填材料)和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离.参照国际上该领域的研究成果,结合我国处置概念,本文就高放废物地质处置中的工程材料(废物固化体、废物罐、外包装、缓冲材料、回填材料),以及其材料选择、设计要求和研究重点等进行了总结.     

黄土高原处置高放废物的初步可行性论证

通过对陕甘黄土高原的地质调查、老黄土和红粘土岩的物理力学性质试验、渗透系数测定、热学特性试验及核素吸附和滞留性能试验,论证了黄土高原处置高放废物的可行性,并对今后深入研究提出建议。研究结果表明在红粘土岩层中或老黄土层中处置高放废物,比国外正在研究的深岩层和粘土层处置具有较大优越性。     

我国高放废物地质处置研究

文章提出我国高放废物地质处置拟采用处置库选址和场址评价-特定场址地下实验室-处置库“三步曲”式技术路线。计划目标是于2030∽2040年前后建成我国的高放废物地质处置库。处置对象是玻璃固化块、超铀废物和部分乏燃料，处置库为竖井-坑道型，候选围岩为花岗岩，位于饱和带中。已初步选定甘肃北山地区为重点预选区。该区地处戈壁，地壳稳定，人烟稀少，地质条件和水文地质条件有利。现已试验获取预选区大量深部地质环境参数。确定使用膨润土作为处置库的回填材料，已获得一批放射性核素在花岗岩和膨润土中的吸附、扩散数据，建立了模拟处置库温度、压力和氧化还原条件的实验装置。高放废物地质处置场址评价、放射性核素地球化学行为、回填材料研究和环境评价研究正在深入进行，并与国际原子能机构等进行了卓有成效的合作。     

中国高放废物处置库缓冲材料选择与基本性能

人类的许多生产、生活活动均可能产生不同活度的放射性废物。其中高放废物由于具有放射性水平高,发热量大,并含有对生物极有害的α放射性的长寿命核素等特点,其安全处置倍受全球科学家和广大公众所重视。目前深地质处置被国际上公认为处置高放废物的最有效可行的方法。借鉴国外成熟的技术和经验,我国采用多重工程屏障系统(包括废物固化体、废物容器及其外包装和缓冲/回填材料)和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离。膨润土由于具有极低的渗透性和优良的核素吸附等性能而被国际上选作缓冲材料的基础材料。经过全国膨润土矿床筛选,高庙子膨润土矿床被选作我国缓冲材料供应基地,我国高放废物深地质处置库缓冲材料的研究以产自该矿床的深部钠基膨润土作为基本组成材料。本文介绍了高庙子膨润土矿床的地质特征以及高庙子钠基膨润土的基本特征。该膨润土与国外同类型材料相比具有蒙脱石含量高(75%左右),杂质矿物相对较少的特点,这对系统和深入研究该材料以开发我国缓冲回填材料技术,确保高放废物的安全有效处置具有重要意义。     

回填材料膨润土性质及其某些工程性能研究

回填材料作为高放废物处置库中的工程阻挡层,填充在废物容器和围岩之间,并可用它封闭废物处置库,密封废物处置区,填满岩石裂缝,对于地下处置系统的安全起着重要的保障作用。回填材料必须具备的特性:(1)具有强吸附放射性核素的能力,阻止和减少放射     

中国高放废物地质处置研究进展:1985～2004

如何安全处置高放废物是核工业可持续发展面临的挑战性问题。我国的高放废物深地质处置研究从1985年开始，提出的计划目标是：于21世纪中叶建成我国高放废物地质处置库，处置的对象是玻璃固化块、超铀废物和部分乏燃料，处置库为竖井一坑道型，候选围岩为花岗岩，位于饱和带中。在1985～2004的20a中，我国高放废物地质处置研究取得了进展，已确定我国高放废物最终处置走“深地质处置”，并且是“三步曲”式的技术路线，即处置库选址和场址评价一地下实验室一处置库。经过全国筛选对比，已初步选定甘肃北山地区为重点预选区，该区地处戈壁，地壳结构完整，地壳稳定，人烟稀少，地质条件和水文地质条件均有利。20世纪90年代初期，开展了地下实验室的选址工作，初步选择了北京郊区2处地点为我国高放废物地质处置“普通地下实验室”的场址。已确定使用膨润土作为处置库的回填材料，并初步确定内蒙古高庙子膨润土为我国高放废物处置库的首选缓冲回填材料。对膨润土的矿物学、岩土力学、物理力学性质和热学性质进行了研究。已获得一批放射性核素(主要是Np、Pu、Tc)在北山花岗岩和膨润土上的吸附分配比、扩散系数和弥散系数等参数，建立了低氧手套箱和模拟处置库温度、压力和氧化一还原条件的小型实验装置。高放废物中的关键核素的化学行为研究也取得进展。花岗岩接触带核素迁移、铀矿床中超铀元素迁移、青铜器腐蚀等天然类比研究取得了成果。还开展了高放废物地质处置系统总性能评价源项和生物圈模式的调研。概念设计研究仅在20世纪90年代初开展了部分研究。从1999年开始，与国际原子能机构开展了2期高放废物地质处置技术合作项目，极大地提高了我们的技术水平。20a的科研工作为我国在21世纪完成高放废物地质处置奠定了一定基础。     

高放废物地质处置库粘土岩场址研究现状

高放废物地质处置库的围岩选择至关重要,粘土岩由于其自身的优点受到越来越多的关注。本文介绍了国际原子能机构（IAEA）、国际经济合作和发展组织（OECD/NEA）等国际组织关于高放废物选址的规范情况,在此基础上介绍了在粘土岩处置研究方面几个有代表性国家如瑞士、法国、比利时的高放废物处置库的选址情况,所选的粘土岩场址以及所选择的粘土岩的特性研究情况。希望可以为我国高放废物粘土岩场址的选择提供借鉴。     

高放废物地质处置容器材料选择和腐蚀行为研究进展

高放废物地质处置容器是高放废物地质处置多重屏障之一,高放废物地质处置容器材料的腐蚀性能决定了处置容器有效性。本文介绍了高放废物容器材料选择的两种策略,以及在不同处置环境下适合的材料,并给出了高放废物地质处置条件下容器材料可能的腐蚀类型。同时介绍了预测高放废物地质处置容器材料寿命的方法和思路,为我国在高放废物地质处置容器的选材上提供新思路和参数。     

中国高放废物地质处置21世纪进展

21世纪近20年,我国高放废物深地质处置进入了一稳步发展的新阶段,在法律法规、技术标准、战略规划、选址和场址评价、工程屏障研究、处置库和地下实验室概念设计、核素迁移和安全评价研究等方面取得了显著进展。其主要亮点包括颁布了《中华人民共和国放射性污染防治法》和《中华人民共和国核安全法》,制定了《高放废物地质处置研究开发规划指南》,颁布了《高放废物地质处置设施选址》核安全导则,确定了2020年前开工建设地下实验室、2050年建成高放废物处置库的目标,甘肃北山预选区被确定为我国高放废物地质处置库首选预选区,建立了场址评价方法技术体系,确定了内蒙古高庙子膨润土为我国高放废物处置库的首选缓冲回填材料,建立了我国首台缓冲回填材料热 水-力-化学耦合条件下特性研究大型实验台架（China-Mock-Up）,获得了一批关键放射性核素的迁移行为数据,开展了初步的安全评价,完成了地下实验室安全技术研究。确定甘肃北山的新场为我国高放废物地质处置地下实验室的场址。2019年5月6日,国家国防科工局批复中国北山高放废物地质处置地下实验室工程建设立项建议书,标志着我国高放废物地质处置正式进入地下实验室阶段。这一系列工作进展和取得的成绩为我国2020年开工建设地下实验室、掌握高放废物地质处置技术奠定了坚实的基础。     

一些国家高放废物地质处置安全评价介绍

本文介绍了高放废物地质处置的安全目标及其准则，并介绍了比利时ＢＯＯＭ粘土高放废物处置性能和安全评价，以及美国玄武岩高放废物处置安全评价的方法和瑞典花岗岩高放废物ＷＰＣ处置方案的安全分析。对我国高放废物地质处置安全评价工作提出了一些建议     

硼铝复合材料热中子屏蔽性能分析

硼铝复合材料因制备工艺简单,力学性能良好,原材料价格低廉等诸多优点被广泛研究,并被用作诸多领域的热中子吸收材料。本文采用理论计算、MCNP软件模拟、实验测量等多种方法对硼铝复合材料的热中子屏蔽性能进行了评估分析。通过理论计算发现,对于相同配比的硼铝复合材料,从材料的热中子吸收性能方面,添加硼单质的效果优于添加碳化硼。通过MCNP程序模拟计算和实验测量发现,硼铝复合材料对能量低于10^-7 MeV的中子吸收效果比较显著。     

Thermal shock response of deformed and recrystallised tungsten

The thermal shock response of tungsten as a plasma facing material (PFM) strongly depends on its mechanical properties and consequently on its microstructure. In order to characterise this influence, deformed tungsten, both in its stress relieved and recrystallised condition, was exposed to 100 ELM like thermal shock events in the electron beam facility JUDITH 1. The induced thermal shock damages were analysed by scanning electron microscopy, optical microscopy and laser profilometry. Tensile tests at different temperatures show that the mechanical properties such as fracture strength and strain depend on the grain orientation and microstructure. Transmission electron microscope images of the as received and the recrystallised material show that the defect density of the recrystallised samples is decreased. Threshold values such as damage and cracking threshold vary with microstructure by a factor of 2. Also the induced thermal shock damages and surface modifications are strongly depend on the microstructure. Surface roughening due to plastic deformation is more pronounced in the recrystallised state and crack parameters as well as crack propagation is influenced by grain orientation due to preferential crack formation along grain boundaries.     

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness J_IC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.     

Irradiation effects on CTR first wall materials

The selection of the controlled thermonuclear reactor (CTR) first wall material (refractory metallic alloy or austenitic stainless steel) will be a compromise based on a number of important nuclear, physical, thermal and mechanical properties, and safety, environmental, technological and economic factors. The correlations between helium production, irradiation, and helium embrittlement in the first wall depend mainly on the neutron spectrum, neutron fluence, irradiation temperature, wall material, reactor operating conditions, etc. The suppression of irradiation swelling by alloying elements (C, Si, Mo, and P) is effective and beneficial. The extent of change in mechanical properties of the first wall material due to neutron irradiation, thermal instabilities (thermal shock, thermal fatigue, crack initiation and propagation), corrosion effect, etc. remains undefined or not completely understood. A fabricated CTR first wall material must meet the requirements of design safety, weldment reliability and good operation performance.     

Study on Solid Self-Lubricating Material for Plasma Facing Components in EAST

In this study, the friction performance of self-lubricating material with the counterpart steel ball-plate rubbing was investigated in vacuum conditions and the thermal distortion of the heat sink sample was tested. The analysis and test results show that the self-lubricating material has excellent anti-friction properties in high vacuum condition and can decrease the thermal stress and avoid damage to the PFCs during physical experiments.     

Radiation effects in thermal properties of polymers. An analytical review. I. Polyethylene

The information about radiation effects on the thermal properties of polymeric materials is of special interest since: (1) the temperature regime of technical structural units operated in the radiation fields is determined in many respects by the values of the material thermal constants and the rate of their degradation; (2) the temperature and dose dependencies of heat conductivity and, especially, heat capacity characterize the polymer structure. The data on radiation effects on the thermal conductivity, heat capacity and diffusivity, linear thermal expansion, density, crystallinity degree, heat and temperature of phase transitions are presented in this review only for polyethylene of high and low density, as a very famous and typical semi-crystalline polymer. It is supposed to deal in the future with such problems as reversible radiation effects on heat capacity, analytical models of the radiation-induced changes and anisotropy effects on the thermal conductivity, accumulated radiation energy (“heat defect”) and an important problem of specific action of different kinds of radiation on the thermal properties of polymers.     

聚氨酯泡沫塑料的辐射效应研究

用DTG和万能材料试验机等分析测试手段研究了聚氨酯泡沫塑料在辐照前后的热稳定性和力学性能变化 ,并利用GC对辐解气态产物进行了定性定量分析。结果表明 ,聚醚型聚氨酯泡沫塑料在 8.0× 10 5Gy时综合力学性能仍然较好 ,但受辐射降解的作用已有相当的气态产物生成。经辐照以后 ,材料的热性能较为稳定     

Lead-lithium eutectic material database for nuclear fusion technology

Fully validated material databases are needed for coherent technological developments in any R&D field. For nuclear fusion technology (NFT), within a near-term perspective of qualification and licensing of nuclear components and systems, this goal is both compulsory and urgent. This mandatory requirement applies for the particular case of the Pb-Li eutectic database as fusion reactor material. Pb16Li is today a reference breeder material in diverse fusion R&D programs worldwide. Technical consensus on most part of the material database inputs seems a major technological objective. In this work Pb16Li material database inputs for NFT have been systematically reviewed. Database inputs (bulk, thermal, physical-chemistry properties, and H-isotopes transport) are discussed and extended to base magnetohydrodynamic (MHD) properties, values for non-dimensional parameters and pipe/channel correlations in 2-phases dispersion models. Ongoing efforts to develop the Pb16Li material database as a computing expert system are reported.     

显微硬度法分析核电站主管道热老化趋势

使用显微维氏硬度计和冲击试验机研究了核电站主管道材料Z3CN20.09M在400 ℃加速热老化10 000 h前后的力学性能变化。结果表明,热老化导致试验材料的冲击吸收能下降;构成试验材料的铁素体相的显微维氏硬度上升,奥氏体相的显微维氏硬度基本保持不变。通过研究材料组织特征,剖析显微硬度与冲击韧性的关系,探索将显微硬度测试方法作为核电站主管道材料热老化趋势预测方法的可能性。     

Use of Alcan's Al-B4C metal matrix composites as neutron absorber material in TN International's transportation and storage casks

Abstract

Major issues in the area of transportation and/or storage of radioactive materials are reliability and safety of engineering components. Among the functions to be undertaken, transportation and storage systems shall allow the criticality control of the transported matter, the control of its temperature, as well as the capacity to withstand the mechanical stresses due to normal, incidental and accidental conditions of use. In most cases, criticality control requires the use of an internal arrangement made of a neutron absorber material, which must also have high thermal conductivity properties to ensure the temperature control. When, as in many AREVA-TN International applications, the design takes credit of the neutron absorber material as a structural component, it must show high mechanical performance. Alcan's Al-B₄C metal matrix composites (Al-B₄C MMCs) meet all the above mentioned requirements, due to their special capability of capturing neutrons, their light weight, and their superior thermal conductivity and mechanical properties. The significant advantage of Alcan's technology is its flexibility with regards to a wide range of boron carbide contents and matrix alloys (from AA1XXX to AA6XXX). This enables the adjustment of the properties to the exact needs of the design. TN International presently uses extruded and/or rolled Al-B₄C MMC parts in several of its internal arrangements. The present paper gives an overview of the manufacture processes of Alcan's Al-B₄C MMCs, from the mixing of B₄C into liquid aluminium to the extrusion and rolling operations. It describes the methods and results for the qualification tests in terms of the neutron absorption, thermal, physical and mechanical properties of the material. Finally, details are given on the use of Alcan's MMCs as a neutron absorber with enough credit for structural material in TN International's TN24 designs.