自蔓延高温合成法固化高放核废料的研究

采用新型氧化剂CrO3,自蔓延高温合成了含Sr2 的钙钛矿.分析了自蔓延高温反应的理论基础,并以XRD、SEM和电感耦合等离子体原子发射光谱仪,分析合成产物的物相、显微结构及浸出率.结果表明:CrO3可将反应燃烧温度由常规1900K提高至2945K;当SrO含量为10wt%时,可以获得较好的钙钛矿物相;当SrO含量为20wt%时可获最低浸出率2.8×10-3g/m2·d.此外,不同配比固化体浸出率皆小于0.13×10-1g/m2·d.     

不饱和聚酯固化模拟放射性废物的可行性研究

本文介绍了用市售191~#不饱和聚酯固化模拟含~(90)Sr 和~(137)C_s 的硝酸钠、偏硼酸钠干盐粉和废湿离子交换树脂的配方、工艺条件和固化体性能测试结果。试验结果表明,不饱和聚酯固化上述放射性废物工艺简单、操作方便,固化体对废物的包容量达50—60%(wt);废物固化体坚硬、密实,外表光滑、平整,具有良好的物理性能及抗浸出性能。     

高放废液碱矿渣水泥固化体研究

介绍了用碱矿渣水泥固化高放废液新工艺。该工艺以碱矿渣水泥为基体,掺加适量沸石和硅灰,无需加温加压,利用静态爆破剂水化时产生的膨胀压,在限容下使水泥基体致密。其抗压强度、耐热性和浸出率等性能均优于其它水泥。当废物氧化物包容量为25％时,固化体抗压强度可达65～100MPa,孔隙率可小于10％,Cs和Sr的浸出率可分别达到10~(-5)g·cm~(-2)·d~(-1)与10~(-7)g·cm~(-2)·d~(-1)水平。高放废液碱矿渣水泥固化体与玻璃固化体性能相当;而固化工艺比玻璃法简单。此外,还探讨了固化机理,核素离子在碱矿渣水泥固化体中的形态。     

模拟高放玻璃固化体的抗浸出性研究

本文对高放废物氧化物包容量为20wt％(含0.131mol/lSO_4~(2-))的GP-12/9B和GP-12/7B两种模拟高放玻璃固化体进行了ISO和MCC-1静态浸泡试验与Soxhlet动态浸泡试验,测定了元素浸出率和质量损失,研究了浸出率与浸泡温度和时间的关系。     

高硫高钠高放废液玻璃固化的配方验证

针对中核四川环保工程有限责任公司(SEPEC)的高放废液高硫高钠的特点,研制出废物包容量为16%(质量分数,下同)的玻璃固化配方。本验证实验研究重点是:验证此配方在改变废物包容量的情况下固化模拟高放废液时对硫的包容能力,以及产生的玻璃固化体的性能是否满足行业标准要求。验证实验中废物包容量分别为12%、13%、14%、15%、16%,玻璃熔制采用了搅拌和不搅拌两种方式。实验结果表明:在不搅拌条件下所熔制的玻璃对硫的包容能力较低,搅拌条件下所熔制的玻璃对硫的包容能力高,但两种条件制备的玻璃的28d总失重、元素归一化浸出率、密度、均匀性、析晶率均达到了我国行业标准的要求。     

模拟低中放废物水泥固化体在地下水中浸出性能的研究

本文采用国家标准方法 ,对硝酸钠、偏硼酸钠和阳离子交换树脂三种模拟低中放废物水泥固化体在浅地层处置环境地下水中的浸出性能进行了实验研究。实验结果表明 ,浸出实验第 42天时 ,硝酸钠、偏硼酸钠和阳离子交换树脂三种模拟废物水泥固化体中 90 Sr的浸出率分别为 1 .5× 1 0 - 6、8.9× 1 0 - 6和 6 .5× 1 0 - 6cm/d,1 37Cs的浸出率分别为 4.5× 1 0 - 6、1 .3× 1 0 - 6和 1 .4× 1 0 - 5cm/d,均优于标准要求。按浸出结果计算了三种模拟废物固化体中 90 Sr和 1 37Cs的扩散系数 ,可作为低中放废物浅地层处置安全评价研究的源项参数     

SiO2和B2O3含量对模拟高放废物玻璃结构的影响

硼硅酸盐玻璃具有特别高的化学稳定性、较好的热稳定性、较大的放射性废物包容量等优点,被广泛作为固化高放废液的基础玻璃料.废物玻璃固化体的结构与其组成存在一定的内在依存关系,它将对废物玻璃的性质产生影响.以组分含量作为变量,所引起的废物玻璃固化体的某些结构特征的变化,是探索影响高放废物玻璃固化体性能的内在线索.为了获取废物...     

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅱ）

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅱ）吴兆广，罗上庚，于承泽，盛嘉伟（中国原子能科学研究院）柳得橹（北京科技大学）关键词高放废物，玻璃固化体，浸出试验，表面分析１引言为了预测高放废物长期处置的安全性，需要研究玻璃固化体的浸出过程和浸出...     

碱矿渣水泥固化模拟高放废液的研究

研究了用碱矿渣水泥固化模拟高放废液。结果表明,碱矿渣水泥在抗压强度、孔结构、Cs离子浸出性能及热稳定性等方面均优于硅酸盐水泥和高铝水泥。以碱矿渣水泥为基体,掺入适量沸石和硅灰,采用特殊工艺,在废物包容量小于25％时。固化体抗压强度65-100MPa,孔隙率小于10％,核素Cs和Sr离子浸出率仅为10~(-5)和10~(-6)g·cm~(-2)·d~(-1)的水平,与现有玻璃固化体性能相近。另外还探讨了核素离子在碱矿渣水泥固化体中的固化机理。     

模拟核素固化体SrZr_xTi_(1-x)O_3的化学稳定性

为研究90Sr衰变子体对90Sr Ti O3固化体结构稳定性的影响,以C4H6O4Sr·0.5H2O、Ti O2及Zr OCl2·8H2O粉体作为原料,按照化学计量比Sr ZrxTi1-xO3(0≤x1)设计配方,采用高温固相法制备一系列固化体。利用X射线衍射光谱(X-ray Diffraction,XRD)、扫描电镜(Scanning Electron Microscope,SEM)及红外(Fourier Transform Infrared,FT-IR)谱仪对制备固化体的物相、结构和微观形貌进行分析表征,并对其抗浸出性能和辐照性能进行了研究。结果表明,当x0.5时,固化体为单一物相;当x≥0.5时,固化体中出现第二相。固化体中的Sr2+、Zr4+的浸出浓度随浸泡时间而增大,在浸泡42 d时,Sr2+的最大浸出浓度为0.006 5μg?m L-1,Zr4+的最大浸出浓度为0.01μg?m L-1。     

Tc immobilization by using perovskite-rich and rutile-rich synroc

1IntroductionSynrocwasinitiallydevelopedin1978byRingwoodetal.[1]Itisasuitableandeffectivemediumforimmobilizingradioactivewaste.Synrochassupe-riorchemicaldurabilitytoconventionalglasswasteforms.Uptonowalotofsynrocapproacheshavebeendeveloped.[2]Tcispresentedinthehighlevelwaste(HLW)withlargerproportion.BecauseTcisaverylong-lived(T1/2=2.14×105a)nuclideitssafedisposalisasignifi-cantproblemconcerned.TccanbeseparatedfromHLWbythepartitioningmethod,thenprocessedbyusingimmobilizationtechnique.Syn…     

Radiation effects of pyrochlore-rich synroc by heavy-ion irradiation

Heavy-ion irradiation is commonly used to study radiation damage of high level radioactive waste (HLW) forms, but S ion was never used before. In this investigation, 100 MeV ^32S ions produced by tandem accelerator was used to study radiation effects on pyrochlore-rich synroc which contained simulated actinides. The amorphization and amorphous doses were determined by X-ray diffractometer (XRD) and transmission electron microscopy/select area electron diffraction (TEM/SAED). The vacancy defects induced by heavy-ion irradiation were characterized by using positron annihilation technique (PAT). The experimental results show that the amorphous dose is 0.5 dpa, the defects produced by heavy-ion irradiation are mainly voids, and irradiation could continue to intensify the vacancy defects even after the amorphous dose was reached.     

Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

A thermal model is constructed and analyses are performed for an ‘in-floor’ type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels.     

Mechanical properties of nuclear waste glasses

The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young’s modulus, Poisson’s ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.     

坑式废物库高水平放射性固体废物回取整备

某核技术利用废库（城市废物库）超期服役,设备老化,20世纪70年代因管理不规范而收贮的高放固体废物给周围环境带来极大的安全隐患,需尽快妥善回取、处置。中国原子能科学研究院放射性三废设施治理工程部根据现场实际工况,设计制订了利用“机器人-气帐”系统回取高放废物的工艺路线。通过现场污染水平和库坑“热点”检测分析,查找到了17号库坑内高放固体废物的具体位置。通过回取实施前的模拟验证试验工作,优化了回取的技术路线。采用远距离回取技术,成功地实施了高放固体废物的回取和包装整备工作。     

膨润土-砂混合物缓冲层轴向热传导性研究

膨润土-砂混合物作为高放废物处置库缓冲材料,在高放废物衰变释热作用下,其物理力学性能对处置库的稳定和安全性具有重要影响。本研究采用自行设计的装置,对按比例缩小后的不同干密度、含水率、掺砂率试样进行热传导模拟试验,并对缓冲层热-力耦合过程进行数值模拟分析,得到了缓冲层温度、应力和应变的变化及分布情况,重点分析了温度的影响。结果表明,增大试样干密度、含水率和掺砂率均可提高其导热性,应变也随之增大,应力受温度影响较早达到平衡;缓冲层靠近热源的位置温度、应力和应变最大,沿轴向方向递减,初始时刻变化明显。     

高放废物处置库雅满苏和天湖地段地下水化

新疆雅满苏和天湖地区是我国高放废物处置库场址的主要预选地段之一。在高放废物处置库选址和评价中,水文地球化学特征是最重要的评价因素之一。在2009-2013年间,开展了雅满苏和天湖地段区域水文地球化学调查,采集地下水样品30余组,根据样品测试结果,讨论了该地区水文地球化学特征,其主要结论是：研究区地下水主要为高矿化的咸水。地下水的pH=7.69～9.64;水温11.8～24.0℃;地下水矿化度（TDS）在542.7～413375.5 mg/L之间变化,地下水的主要化学类型为Cl•SO₄-Na、SO₄•Cl-Na和Cl-Na型。控制地下水化学形成的主要作用是溶滤作用和蒸发浓缩作用。     

玻璃固化过程中Mo的化学行为研究进展

高放废液中的Mo在玻璃中的溶解度较低,在高放废液的玻璃固化中易形成黄相,黄相的出现对玻璃固化过程和固化体性能均不利,限制了玻璃固化体中废物的包容量。通过改变玻璃配方或开发研究新的配方提高固化体中Mo含量,可以消除黄相。本文综述了近年来国内外针对玻璃固化过程中Mo的化学行为研究所取得的研究进展。     

Solidification of simulated actinides by natural zircon

Natural zircon was used as precursor material to produce a zircon waste form bearing 20wt% simulated actinides (Nd2O3 and UO2) through a solid state reaction by a typical synroc fabrication process. The fabricated zircon waste form has relatively good physical properties (density 5.09g/cm^3, open porosity 4.0%, Vickers hardness 715kg/mm^2). The XRD, SEM/EDS and TEM/EDS analyses indicate that there are zircon phases containing waste elements formed through the reaction. The chemical durability and radiation stability are determined by the MCC-1 method and heavy ion irradiation; the results show that the zircon waste form is highly leach resistance and relatively stable under irradiation (amorphous dose 0.7dpa). From this study, the method of using a natural mineral to solidify radioactive waste has proven to be feasible.     

高放废物处置评价中应用的有关程序

对有关在高放废物处置中应用的程序进行了调研,根据应用方向将这程序分成源顶计算程序、地下水与核素迁移程序、热－湿－力耦合程序及其它程序4类。