核电站废树脂清洁解控问题探讨

对轻微污染的废树脂采取解控处理是核电站废物最小化工作的重要内容。在介绍近几年开展的核电站废树脂解控监测工作的基础上,对核电站目前的废树脂废物管理中存在的问题进行了总结,提出了有利于实施废树脂废物最小化的技术途径。解控后的废树脂可能采取的处置出路包括危险固体废物填埋处置和水泥窑协同处置,暂时尚不具备进行高温焚烧处置的条件。为了提高核电站这类废物最小化的实施效率,建议对不同类型放射性核素在废树脂中含量的相关性问题,对放射性核素在离子交换柱树脂中的空间分布问题,对废树脂分拣的在线监测技术以及其它类型固体废物的分拣监测问题等进行深入研究,以期在保证取样代表性的同时,减少取样量和测量工作量,提高工作效率,降低废物最小化的经济成本。     

废放射源近地表处置活度限值推导

以我国低中放固体废物北龙处置场和西北处置场为参考场址,针对钻探景象和钻探后景象,考虑了几个重要放射性核素,推导了废放射源在两个场址中处置的活度限值。除¹³⁷Cs的关键景象为钻探景象外,其他所选核素的关键景象均为钻探后景象。且单个废物包中的废放射源活度限值不受处置场的限制,推荐了适合于近地表处置的单个废物包内废放射源的接收限值,⁹⁰Sr、¹³⁷Cs、²³⁸Pu、^63Ni、²⁴¹Am、²²⁶Ra和²³⁹Pu分别为2.0×10¹⁰、6.0×10⁹、1.0×10⁶、1.0×10¹¹、1.0×10⁷、1.0×10⁶和8.0×10⁶Bq。     

秦山核电基地放射性废物最小化技术实践与探讨

介绍了秦山核电基地在放射性废物最小化方面的实践。包括建立了蒸汽发生器排污树脂和通风过滤器金属框架清洁解控工作流程;改进了水泥固化线工艺,水泥固化包装容器采用金属桶代替水泥桶;开发了废过滤器芯子暂存衰变法分类处理及一桶装多芯方案;开发了技术废物“三明治”式废物装桶和超级压实提高外包装容器效率等最小化实践。对废液、废树脂、技术废物、有机废液等处理技术进行了比较分析,建议采用废液蒸干压实技术、废树脂蒸汽重整技术、可燃废物和有机废液焚烧技术,预计秦山核电基地废物产生量可由541 m³/a减至约94 m³/a。     

Chapter 3: Type and Quantity of Radioactive Waste Destined for Disposal

Abstract

According to the current planning status, radioactive waste with negligible heat generation is destined for storage in the Konrad waste repository in accordance with the Provisional Waste Acceptance Requirements⁽⁴⁾. Waste of this kind occurs in the nuclear fuel cycle, research, medicine and technology. In its original state, the primary wastes of this type have various forms, such as: liquids, concentrates, sludges; ion exchange resins; compressible and/or combustible solids; incompressible solids, e.g. structural material components; filters, filter candles; other types of waste. The radioactive waste is appropriately conditioned and packaged before transport to the waste disposal site. To meet the basic requirements specified in the Waste Acceptance Requirements, primary waste must be solidified, for which the most common solidifying agents are cement and concrete although bitumen is also used. Waste products in decomposing, fermenting or liquid form or which contain a significant fraction in such states are not accepted for disposal.     

低中水平放射性固体废物近地表处置设施安全分析报告格式和内容的探讨

我国目前尚缺少放射性废物处置安全分析报告格式、内容要求的相关导则,本文在基于我国已有低中放固体废物近地表处置工程实践及国际原子能机构安全全过程系统分析理念的基础上,对近地表处置安全分析报告的编制要求进行探讨,提出对应的格式与内容建议,以期为编制低中水平放射性固体废物近地表处置设施安全分析报告格式和内容给的标准提供参考。     

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).     

Demonstration of structural performance of IP-2 packages by advanced analytical simulation and full-scale drop test

Two new types of IP-2 (Industrial Package Type 2) to transport low and intermediate level radioactive waste (LILW) steel drums from nuclear power plants to a disposal facility have been developed in accordance with the IAEA and Korean regulations for radioactive materials. According to the regulations, both packages must preserve their structural performance after they are subjected to 0.9 m free drop tests, which are prescribed as normal conditions.In this study, an advanced analytical simulation and an evaluation process using the finite element (FE) method have been developed for the design assessment of the newly developed IP-2s. Then, analytical simulations for the various drop orientations were performed to evaluate the structural performance of the packages and demonstrate their compliance with the regulatory requirements. Also, full-scale drop tests were carried out to verify the numerical tools and modeling methodology used in the analyses and to confirm the performance of the IP-2s. In addition, parametric studies are carried out to investigate the sensitivity of the analytical variables, such as the material model and modeling methodology.In addition, this paper intends to provide basic guidance on the analytical simulation and evaluation process specifically for Korean types of transport packages, because numerous transport packages must now be developed for the various kinds of LILW that have accumulated in temporary storage facilities in Korea.     

固体放射性废物水泥砂浆固定配方工程验证试验

为完成某废物库回取、整备设施中整备车间固体废物的整备,满足车间水泥固定系统的工程应用需求,对初期实验室研究的中低放固体废物固定配方在该系统上进行200 L废物桶和2 m³废物箱的工程验证试验,考察该配方的固化体性能,如固化过程中的温升、固化体均匀性等,并确定加料顺序和搅拌时间等工艺参数。结果表明,生产的固化体符合行业标准(EJ 1186-2005)的要求,满足国家废物运输和处置要求。     

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.     

放射性固体废物水泥砂浆固定配方研究

本工作主要研究中、低放固体废物超级压缩饼在2m3废物包装箱内的固定配方。该配方能够保证最终废物体的整体性和整体强度满足安全运输、储存和处置的要求,并且能够进行工程应用。通过实验研究水灰比、灰砂比、砂子级配以及添加剂加入量等因素对流动度、凝结时间、固化过程中的温升、固化体性能等的影响规律,根据规律筛选出既满足核行业标准(EJ1186-2005)又适用于现有工程装置的放射性固体废物水泥砂浆固定配方,即在室温25℃,常压下,配方为0.450∶1水灰比、1∶1.6灰砂比、1∶2∶1粗中细砂比、1‰(质量分数)B型缓凝剂。     

环氧树脂固化放射性废树脂的配方初探北大核心CSCD

为了将中、低放射性废树脂固化成稳定的固化体,采用环氧树脂为基材,通过添加合适的固化剂、稀释剂、阻燃剂等进行固化配方试验研究。结果表明,配方为E-44环氧树脂∶651固化剂∶添加剂=1∶0.53∶0.05(38∶20∶2)、质量包容率50%时,其固化体性能最优,并满足国标GB 14569.2的要求,研究结果表明采用环氧树脂体系固化放射性废树脂是可行的。     

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。     

Packaging Design and Qualification: The Experience of the CDTN/CNEN

Abstract

Since 1982 the CDTN, the Nuclear Technology Development Centre, has been designing, testing and qualifying packaging for radioactive materials. These packagings are used for the transport of radioisotopes and disposal of spent sealed sources, wastes generated in the nuclear fuel cycle and the wastes produced in the radiological accident that occurred in the city of Goiânia. For radioactive tracers and medical/industrial radioisotopes, the packagings used are cardboard and wood boxes, while the spent sealed sources are preferably conditioned in metal drums containing lead shielding and a gas absorber material. To condition and transport the wastes from the various nuclear cycle activities, metal drums and boxes are used in Brazil. For the higher active wastes from the nuclear power plant Angra I, a metallic drum in a concrete overpack is used. The wastes generated in the accident were first conditioned in the readily available packaging. Later on, more appropriate packaging was designed by the CDTN staff. CDTN has carried out a programme since 1983 to evaluate the durability of commercial drums used for waste conditioning.     

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

放射性废树脂氧化分解处理技术

简要介绍了放射性废离子交换树脂（以下简称废树脂）的来源、特点和传统处理方法。概述国内废树脂氧化分解处理技术及其进展,重点论述湿法氧化分解的工艺原理、反应机理以及工艺流程的发展现状。分析评述废树脂分解液水泥固化技术研究的最新动态及成果。     

核电厂废过滤器芯子分类处理方法与实践

废过滤器芯子是核电厂放射性固体废物的主要来源之一,为减少最终废物产生量,根据表面剂量率水平进行分类处理。表面剂量率2 mSv/h以下的废过滤器芯子进行烘干、压实、灌浆固定;表面剂量率在2～50 mSv/h的废过滤器芯子暂存衰变,然后再烘干、压实、灌浆固定;表面剂量率大于50 mSv/h的废过滤器芯子直接灌浆固定。按照上述方法分类处理废过滤器芯子,以设计产生量估算,两台机组每年可以减少最终的废物包21.3 m³;以实际产生量估算,两台机组每年可以减少最终废物包8 m³。废过滤器芯子分类处理符合放射性废物最小化管理原则,可有效减少废物的处理和处置费用,对核电发展具有明显的经济效益和环境效益。     

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。     

Effect of borate concentration on solidification of radioactive wastes by different cements

To investigate the effect of borate on the cementation of radioactive evaporator concentrates, and to provide more data for solidification formula optimization, the simulated borate evaporator concentrates with different borate concentrations (as B) and Na/B ratio (molar ratio) were solidified by sulfoaluminate cement (SAC) and Portland cement (PC), with addition of Ca(OH)₂, zeolite and accelerator or water reducer. The hydration products of solidified matrices were characterized by X-ray diffraction (XRD). The experimental results showed that borate retarded the cement setting for both SAC and PC formulas, and the final setting time prolonged with decrease of Na/B ratio. Borate could enhance the fluidity of the cement mixture. The 28 d compressive strengths of the solidified matrices for both SAC and PC formulas decreased with increase of borate concentration. The XRD patterns suggested that, in the matrices maintained for 28 d, borate did not interfere with the formation of main hydration products of SAC and PC. Borate, in the form of B(OH)⁴⁻, incorporated in ettringite (3CaO·Al₂O₃·3CaSO₄·32H₂O) as solid solution phase. The formula of SAC and PC developed in this study was effective for cementation of the simulated borate evaporator concentrates. However further optimization was required to reduce retarding effect of higher borate concentrations and to extend the practical feasibility for actual evaporator concentrates.     

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.