水泥浆固化体的浸出行为

本文概述了用于水力压裂处置的水泥浆固化体在去离子水和处置现场近域地下水中的浸出行为。浸出试验分别采用国际标准化组织(ISO)颁布的长期浸出试验方法和美国能源部材料特征化中心(MCC)曾采用过的静态浸出试验法。用长期浸出试验方法(浸出剂为去离子水)得到192天时的三种浸出试样(Ⅰ、Ⅱ、Ⅲ)中~(137)Cs 和~(90)Sr 的浸出分数、累积浸出分数、递增浸出率和扩散系数见下表。     

放射性废物固化体抗浸出性快速测定方法探讨

固化体的抗浸出性是放射性废物安全管理的一重要参数。目前，国内采用国标GB7023—86中的标准浸出试验方法测试固化体的抗浸出性，试验周期长。并且，国标GB14569.1—93仅对核素第42d的浸出率作了规定。这一规定不能很好反映不同固化基材、不同配方固化体间抗浸出性的差异。美国国家标准ANSI/ANS-16.1—2003采用快速浸出试验方法，并用浸出因子来表征核素的抗浸出性。本工作参照美国标准对试验结果的处理方法，对以往获得的真实或模拟放射性废物水泥固化体的浸出试验数据进行重新计算。计算结果表明，当核素累积浸出百分数小于20%时，核素的浸出率与浸出因子间存在一定的换算关系。据此，可考虑建立快速浸出试验方法和新的试验结果表述式，以较全面地判定放射性废物固化体的抗浸出性能。     

固化体长期浸出实验的改进

浸出试验是测定放射性废物固化体性能的主要手段之一.IAEA于1969年推荐了一个标准浸出试验方法,其中对浸出实验的有关方面都做了规定.不少从事浸出试验的工作者都参考这个方法,同时又根据各自的情况进行了必要的修改.由于IAEA推荐方法本身就存在一些缺陷,再加之不同的修改,致使各工作者测定结果的相比性减弱了,有些甚至无法相比较.ISO于1982年发布了ISO6961,此国际标准也有一些不妥之处.为制定我国的标准浸出试验方法,有必要对浸出试验的细节进行试验研究和验证.影响浸出试验准确度的因素很多,因篇幅所限,本文只能对部分问题进行探讨.     

关于报告快速浸出试验结果的建议

本文提出报告快速浸出试验结果的建议,即用与(?)相对应的累积浸出时间、累积浸出分数及递增浸出率来报告试验结果。这些数据可用于筛选固化工艺配方、评价固化工艺及预测同化体中放射性核素的释出量等。     

关于中低放废物固化体浸出试验中几个问题的讨论

本文就中低放废物固化体浸出试验中的V_L/S_A值、浸出试样体积的下限值,浸出液的分析和浸出试验的终止时间的判断等问题进行了讨论。     

测定放射性废物固化制品浸出因子的建议

本文对1971年国际原子能机构(IAEA)所推荐的放射性废物固化制品的浸出率试验方法提出一个改进意见。建议用圆柱形试样、不加密封直接浸泡的方法进行浸出率试验,以简化试样的制备步骤,加速试验进程和改善测量精度。在简单回顾了 IAEA 法的理论基础以后,给出了圆柱形固化块扩散方程的求解方法和实用的计算曲线,从而使这种圆柱形试样的浸出试验数据也能用浸出因子 L(或扩散系数 D)来表示。利用 L 可将不同实验室来的、在试验细节上有一定差别的试验结果进行相互比较。     

放射性废物形式浸出性研究近况

前言贝恩斯(R.N.Burns)在“低、中放废物的固化”一文中曾对1969年以前的固化体浸出性研究情况做过概述;门德尔(J.E.Mendel)在为美国“废物固化开发工作组”所管辖的“浸出性试验方法小组委员会”提供有关浸泡试验方法和结果的现状而写的一篇题为“浸出试验方法和各种含放射性废物的固体介质浸出性评述”的调研报告中曾对1973年以前的浸出性研究成果进行了详尽的总结和评述;笔者在1980年前后对1973年以     

铀矿石的硫酸高铁浸出研究

利用硫酸高铁浸出剂对中国湖南省南部某铀矿进行了浸出研究,该矿铀矿物主要为沥青铀矿,试验证明,高铁离子浸出比常规硫酸浸出省酸35～40％,矿物杂质浸出量减少15～20％。借助铁氧化细菌将Fe~(3 )浸出剂循环使用,6次循环浸出结果证明,铀的浸出效果未发生不良变化。     

铀矿石的硫酸高铁浸出研究

利用硫酸高铁浸出剂对中国湖南省南部某铀矿进行了浸出研究,该矿铀矿物主要为沥青铀矿,试验证明,高铁离子浸出比常规硫酸浸出省酸35～40％,矿物杂质浸出量减少15～20％。借助铁氧化细菌将Fe~(3 )浸出剂循环使用,6次循环浸出结果证明,铀的浸出效果未发生不良变化。     

巴基斯坦地浸采铀的野外试验

在巴基斯坦西北部的班努地区,圈定了一些产于西瓦里克群中的小型铀矿体。含矿主砂岩很松软并且未固结,因而用常规方法难以开采。库布尔克尔矿体符合用原地浸出法开采的大部分条件。矿石的实验室浸出试验世得出了令人鼓舞的结果。因而采用碳酸钠和重碳酸钠的混合液作为浸出剂在巴基斯坦进行了地浸采铀的现场试验。试验证明铀矿体可以用碱法浸出;如同时使用过氧化氢做为氧化剂,则浸出率提高;若用氧代替过氧化氢,则生产总成本可大大降低。     

Solidification of spent radioactive organic solvent by sulfoaluminate and Portland cements

The solidification of simulated spent radioactive organic solvent, tri-butyl phosphate/kerosene, was investigated by emulsification–solidification method using sulfoaluminate cement (SAC) and Portland cement (PC). Zeolite, calcium hydroxide and MR-1 type emulsifier were mixed into the cement blends for improving the performance of solidified waste forms (SWF). The properties of SWF were evaluated in terms of mechanical strength, leachability and mineral phase analyses. The hydration products of SWF were characterized by X-ray diffraction (XRD). The experimental results showed that the 28 d compressive strengths of SAC solidified waste forms (SACF) and PC solidified waste forms (PCF) were 14.23 and 19.07 MPa, respectively. Leaching sequence of three radionuclides in two kinds of SWF is Cs⁺ > Sr²⁺ > Co²⁺. Compared with PCF, SACF had better performance in preventing nuclides Co²⁺ and Cs⁺ from leaching to the environment. The XRD patterns suggested that simulated spent radioactive organic solvent and emulsifier in SWF did not obviously change the hydration products of the two cements (SAC and PC).     

模拟放射性含硼废液的水泥固化研究

为了比较硫铝酸盐水泥和普通硅酸盐水泥含硼废液的固化,为配方优化提供依据,研究采用两种配方对模拟放射性含硼废液进行水泥固化。测定了固化体28d抗压强度、抗浸泡性、抗冻融性和耐γ辐照试验后的强度损失,进行了模拟核素浸出试验,并对固化体水化产物进行XRD分析。结果表明,两种配方可有效固化模拟含硼废液,固化体28d抗压强度、各项试验强度损失和模拟核素浸出率均满足GB14569.1—93的要求,试验所用的硫铝酸盐水泥配方对Cs＋的滞留能力优于普通硅酸盐水泥配方,固化体中的硼以B(OH)4－形式固溶在钙矾石中。     

搅拌参数对放射性废液水泥固化体性能的影响

目前,国内核电站或核设施产生的中低放废液都采用水泥固化进行处理,水泥浆及水泥固化体性能是水泥固化技术重点研究内容。本文采用普通硅酸盐水泥固化中低放废液模拟料液,研究不同液灰比条件下,搅拌时间和搅拌速度对水泥浆流动度和固化体28 d抗压强度、孔结构、显微结构和抗浸出性能的影响。结果表明：在相同液灰比下,随着搅拌时间的延长（10～50 min）,水泥浆的流动度和固化体抗压强度呈现先增大后减小的趋势,而固化体的孔隙率和Sr²⁺浸出率随搅拌时间的延长呈递减的趋势,搅拌50 min的固化体的结构较搅拌10 min的固化体致密;用较大搅拌速度制备的固化体的抗压强度较高,且在搅拌30 min内,提高搅拌速度可提高浆料的流动度;然而长时间用较大速度搅拌制备的固化体的孔隙率较高,同时核素浸出率也较大。由于固化工艺过程中搅拌速度和搅拌时间会影响水泥浆的流动性和固化体性能,因此在水泥固化装置投入使用前,应通过大量实验来确定满足工艺要求且满足固化体性能的最佳搅拌参数。     

桶内搅拌水泥固化实验装置

本文描述了放射性废液水泥固化的中试实验装置和固化过程,给出了该装置较好的操作条件。它用模拟泥浆和浓缩液进行了一些有关该装置的性能实验,其结果表明,这个装置是可用的,由制得的固化体的物理性能令人满意。     

Disposal concept for LILW in Korea: Characterization methodology and the disposal priority

Since 1977 the Korea Hydro and Nuclear Power Co. has generated about 67,000 drums (200 L) of low and intermediate level radioactive waste (LILW) and the drums are stored in the temporary storage facility at each reactor site. The accumulated dry active waste (DAW) amounts to around 36,600 drums. There are around 19,000 drums with evaporator bottoms, 9700 drums of spent resin, and 1600 drums of spent filters. This study proposes four mandatory items with regard to the radioactive characterization of LILW: namely, the total activity, surface dose, individual activity, and surface contamination. The required contents of the physical characterization include the weight (density), voidage, free liquid, and homogeneity. For the chemical characterization, the required contents include leachability, corrosiveness, explosiveness, and chelation. Finally, the compressive strength and integrity of drums are requested for the mechanical characterization. To determine the disposal priority of LILW in the Republic of Korea, the authors considered two main factors, namely, the waste management situation in Korea and overseas case studies. After considering those factors, the authors established a disposal priority for the LILW: (1) concentrated waste solidified with the cement and low radioactive DAW, the characterization of which can be readily identified in detail; (2) spent resin solidified with cement; (3) spent filters; (4) highly radioactive DAW, the characterization of which is well documented; and (5) waste that needs to be researched further, including spent resin in PE-HIC, evaporator bottoms and concentrated waste solidified with paraffin, and DAW that contains some harmful materials.     

Utilization of Crushed Radioactive Concrete for Mortar to Fill Waste Container Void Space

Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in a shallow burial disposal facility as low level radioactive waste must be solidified by cement or other materials with adequate strength and must provide no harmful opening. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete for fine aggregate for mortars to fill void space in waste containers. Tests were performed with pre-placed concrete waste and with filling mortar using recycled fine aggregate produced from concrete. It was estimated that the improved method substantially increases the waste fill ratio in waste containers, thereby decreasing the total volume of disposal waste.     

模拟放射性含氟废液水泥固化配方的研究

本文以沸石、硅灰、石英砂为添加剂,按照质量比m（沸石）∶m(硅灰)∶m（石英砂）∶m（水泥）=1∶1∶3∶10配方对模拟放射性含氟废液进行水泥固化。由配方得到的水泥浆流动度和初、终凝时间满足桶内固化要求。测定了水泥固化体28 d的抗压强度、抗浸泡性和抗冻融性实验后的强度损失,进行了抗冲击性能测试和模拟核素浸出实验。结果表明,该配方可有效地固化模拟放射性含氟废液,固化体28 d抗压强度、各项实验强度损失和模拟核素浸出率均满足GB 14569.1-2011的要求。水泥固化体的F^-浸出率很低,XRD显示F^-以CaF₂形式存在。废液中F^-质量分数控制在1%较为合适,此时水泥固化体终凝时间为14 h,F^-的42 d浸出率为2.54×10^-3 cm/d。     

Solidification of borate radioactive resins using sulfoaluminate cement blending with zeolite

The solidification of borate radioactive resins using sulfoaluminate cement (SAC) blending with zeolite was investigated, and the performance of SAC–zeolite matrix for immobilizing borate resin was evaluated based on the compressive strength, leachability and mineral phases of solidified waste forms. The experimental results showed that the 28 d compressive strength of resin SAC–zeolite waste forms was 18.8 MPa, the cemented waste forms meet the requirement of water, freezing, irradiation and impact resistance tests. The leaching processes of Cs⁺ and Co²⁺ were diffusion-limit step. The leachability of Cs⁺ was higher than Co²⁺ no matter what the matrix contained zeolite or not, but leachability of both cations was reduced by adding 5% zeolite. For SAC–zeolite matrix, the leaching rates of Cs⁺ and Co²⁺ were 6.40 × 10⁻⁵ cm/d and 1.62 × 10⁻⁷ cm/d, and the cumulative leaching fractions were 2.9 × 10⁻³ cm and 2.3 × 10⁻⁵ cm, respectively. The X-ray diffraction (XRD) analysis suggested that B(OH)₄⁻ was incorporated by substitution of SO₄⁻ in the crystalline structure of ettringite (3CaO·Al₂O₃·3CaSO₄·32H₂O).     

放射性核素从废物体中向外释放行为初步研究

以去离子水为初始释放环境,考察了1年释放周期内,两类典型低、中水平放射性废物固化体在被水完全浸泡情景下,3种关键核素⁶⁰Co、¹³⁷Cs、⁹⁰Sr的向外释放行为。使用平均有效扩散系数作为表征参数。多个样品的测试结果表明：3种核素向外释放表现出规律性的趋势,即¹³⁷Cs最快,⁶⁰Co最慢, ⁹⁰Sr相比居中。对于不同的核素,起始活度对释放行为的影响是不同的。借助SEM探讨了废物体微观结构对核素释放的影响。