低、中水平放射性废物水泥固化体性能检验的讨论

放射性废物固化体的性能检验是保障放射性废物安全处置的有效措施之一.对于低、中水平放射性废物水泥固化体性能要求和性能检测,有关的国家标准中有明确规定.本文根据我国放射性废物水泥固化工作的实际需要,从引用的标准、抗压强度、抗浸出性和耐γ辐照性4个方面对现行国家标准<低、中水平放射性废物固化体性能要求水泥固化体>需要修订和更新的部分内容进行初步讨论.     

放射性废离子交换树脂水泥固化与机理探讨

针对放射性废离子交换树脂稳定化处理技术现状,研究了适合现阶段我国放射性废离子交换树脂水泥固化的工艺,并利用XAD和SEM分析技术探讨研究了废树脂水泥固化体的结构和性能及采用新型ASC水泥作为固化基材的基本理论依据.     

废树脂苯乙烯固化研究(Ⅱ)

本文研究废树脂苯乙烯固化过程的热效应,辐解气体和热解气体的生成,树脂固化产品对金属材料的腐蚀作用以及固化产品抗老化性能。     

废离子交换树脂的优化处理

核设施产生的废树脂的安全处理、整备和处置是热点问题。本文论述了废树脂的特殊性 ,解析了各种废树脂处理、整备技术 ,包括脱水干燥后装入高整体性容器、洗脱、热压、生物降解、焚烧、湿法氧化、沥青固化、聚合物固化、玻璃固化和水泥固化等。重点分析了废树脂水泥固化 ,讨论了树脂溶胀作用破坏固化体的机理 ,介绍了克服树脂溶胀作用的方法。强调指出必须重视水泥固化的配方 ,关键是必须满足处置要求。最后 ,对优化处理废树脂提出了建议     

放射性废物处理与处置标准在中放废液大体积浇注水泥固化工程中的应用探讨

本文扼要阐述了放射性废物处理与处置标准在中放废洼失体积浇注水泥固化工程中的应用,针对GB 14569.1-93<低、中水平放射性废物固化体性能要求水泥固化体>和GB 7023-86<放射性废物固化体长期浸出试验>放射性废物处理与处置标准在中放废液太体积浇注水泥固化工程应用中所存在的问题进行了探讨,并提出了建议.     

模拟放射性树脂特种水泥固化提高包容量的研究

研究了用ASC特种水泥固化放射性废树脂过程中增加树脂包容量对固化体强度、核素浸出率和水化热温升的影响.研究结果表明增加树脂包容量将使固化体强度有所降低,但当树脂包容量不大于60%的时候仍然可以满足废物处理和处置的要求;增加树脂包容量对核素浸出率影响很小,并且可使固化过程中的水化热温升有所降低.     

放射性废物水泥固化样品核素浸出率测量分析

以放射性废树脂、残渣和蒸残液的水泥固化热配方试验为依据,运用HPGe-γ谱仪、低本底α、β测量仪对废物固化样品的放射性核素浸出率进行测量,分析不同源项的水泥固化体核素浸出率结果,验证相应水泥固化样品配方的准确性及可靠性。结果表明,残渣、蒸残液和废树脂的不同水泥固化样品中60Co、137Cs和总β的浸出率均在浸泡前期急剧下降;随着浸泡时间的延长,浸出率变化趋于稳定;浸出率满足GB14569.1-93的要求。     

聚合物水泥固化放射性废树脂配方

采用环氧树脂乳液与复合水泥制备的聚合物水泥固化模拟放射性废树脂,应用正交设计法进行试验设计。首先进行复合水泥配方的正交设计,确定复合水泥中525#快硬硫铝酸盐水泥、硅粉、沸石和粉煤灰之比为1∶0.05∶0.10∶0.05,然后进行固化树脂配方的正交设计。以抗压强度作为鉴定废物固化体的物性依据,应用F检验,选择优化的配方。最终选择优化的复合水泥作固化基质,环氧树脂乳液作胶凝材料。优化配方为:乳灰比,0.55;树脂包容量,0.3;阴阳树脂比,2∶1。根据GB14569.1—93的要求对采用该优化配方的废物固化体进行了性能测试。结果表明,得到的水泥固化块(50 mm×50 mm)的抗压强度大于10 MPa,固化体的抗冻融、抗浸泡、抗冲击、抗辐照性能满足废物近地表处置的要求。     

放射性废离子交换树脂特种水泥固化体的微观结构分析

研究了特种水泥 (ASC)树脂固化体的微观结构。用压汞实验比较了ASC特种水泥的树脂固化体和普通硅酸盐水泥 (OPC)固化体多孔性能 ,通过电镜扫描 (SEM )观察比较了ASC和OPC的微观晶体结构。分析结果发现ASC水泥固化体具有较好的孔形结构 ,这是ASC固化体浸出率低的原因 ;ASC水泥固化体晶体呈针状结构 ,OPC水泥固化体晶体呈片状结构 ,针状结构的力学性能和结构强度要比OPC的片状结构好 ,该结构是ASC固化放射性废树脂包容量大、强度高的根本原因。     

聚丙烯纤维对废树脂水泥固化体性能的影响

为了避免或降低放射性废树脂水泥固化体因吸水溶胀而开裂的可能性,在原配方的基础上添加了聚丙烯纤维。试验结果表明,纤维材料的掺入可有效限制固化体裂纹的增长、改善固化体的脆性,在一定程度上提高固化体的强度、抗浸泡性及抗冻融性。在水灰比0.35,湿树脂体积包容量40%,聚丙烯纤维体积掺量为0.2%时,固化体抗压强度可以达到20 MPa左右,浸出率与抗水性也均满足有关标准要求。     

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。     

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.     

硼饱和模拟废活性炭的桶外水泥固化

核电站在运行过程中会产生含硼废水浸泡且硼饱和的废活性炭,需要对其进行固化处理。采用2.5‰(质量分数)的硼酸溶液对活性炭进行浸泡直至含硼量达到稳定来模拟实际废物源项,然后脱去硼饱和后活性炭中的游离水,并采用桶外水泥固化工艺对其进行水泥固化。结果表明,试验过程稳定可靠。另外通过试验验证可知,华龙一号新增废物源项废活性炭能够利用现有桶外水泥固化技术进行废物的处理,对硼饱和模拟废活性炭采用桶外工艺固化后,养护形成的水泥固化体按照标准GB 14569.1-2011进行游离液体、机械性能、抗水性、抗冻融性以及耐γ辐照性试验测试,结果均满足要求,且活性炭颗粒在固化体中分布均匀,未出现上浮现象,其搅拌完成以及30min后自由流动度均大于200mm,完全满足桶外水泥固化工艺稳定可靠的运行要求。     

碳钢容器腐蚀模式

碳钢是用于低中放废液贮存或低中放固体废物处置包装容器的一种常用材料。容器的完整性是阻滞废液向外泄漏或者被处置废物中放射性核素向外释放的一个重要因素。包装容器损坏的主要机制是腐蚀。在废液储存或废物处置条件下，所关心的是容器的点蚀和均匀腐蚀。本文介绍了有关这方面的情况和一些点蚀模式和均匀腐蚀模式。     

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.     

Solidification of Spent Ion Exchange Resin Using New Cementitious Material, (II)

When using cement solidification for spent ion exchange resin, resin content in the waste form is typically controlled below 20vol%. This is because the waste forms crack and deteriorate in water at higher resin contents. The deterioration mechanism and its preventive measures were investigated in this study.

Swelling pressure of the resin was experimentally measured for various waste forms. The resin in the waste form tended to swell under a water immersion condition and tensile stress was exerted on the cement matrix. The stress value changed in the region of 1.5~5.5 MP a depending on such factors as resin content and its type. The waste form was deteriorated in water when the tensile stress was higher than the tensile strength of the cement. These results suggested that the deterioration could be prevented by increasing the tensile strength of cement. A fiber reinforced cement was developed for which tensile strength was almost doubled (~7MP a) by adding 10 wt% steel fiber to the cement. This prevented deterioration and allowed the resin content to be increased from 20 to 42 Vol%     

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。     

Reduction of Cesium Leaching Ratio from Cementitious Resin Forms Using Natural Active Silica

Cesium adsorption behavior of active silica, which is a natural acid clay composed of cristobalite and quartz, was evaluated for its applicability as Cs adsorbent to be added to cementitious waste forms containing spent ion exchange resin. Since active silica carried the Cs exchangeable silanol group (—SiOH) originally, the Cs distribution coefficient was remarkably high (<10⁴). It increased in saturated Ca(OH)₂ solution, simulating the cement paste, due to formation of new silanol groups. With its addition to the cementitious forms with ¹³⁴Cs adsorbed ion exchange resin solidified by slag cement, the Cs leaching ratio was reduced to below 1/10 that without active silica.     

放射性污染泥土的整备

²⁴¹Am污染的低放水平泥土,作为一种放射性固体废物,在核设施退役中尚无成熟的整备技术。本文介绍一种比较特别的整备方式,首先将放射性泥土在包装容器（200 L碳钢桶）内采用20 t桶内压缩处理,将放射性泥土从松散状态压缩至密实状态,压实比为1.1～1.4。然后在包装容器上部浇注10～15 cm厚水泥砂浆进行密封,形成以放射性泥土为核心的、可用普通的固体废物描述的水泥密封体。将水泥密封体作为整备对象,采用Ⅶ型钢箱进行再包装和水泥固定整备,最后形成Ⅶ型钢箱包装体。通过有关辐射监测,确定Ⅶ型钢箱包装体放射性特性参数符合放射性废物运输要求,满足国家处置场接收和最终处置的要求。