超临界二氧化碳中TBP络合萃取硝酸钕

将超临界流体萃取技术应用于乏燃料后处理中,可简化后处理流程、减少二次废液的产生。本工作进行了含磷酸三丁酯(TBP)的超临界二氧化碳(SC-CO2)络合萃取硝酸钕的实验研究,考察了硝酸钕粒径、TBP流量、系统温度和压力对络合萃取过程的影响。实验结果表明,含TBP的SC-CO2可高效萃取硝酸钕,萃取率达97%以上,增大TBP流量可加快萃取过程,而粒径、温度和压力对萃取速率的影响则较小。由实验结果可推断,该络合萃取过程的动力学受络合反应控制,并用一动力学模型计算出表观反应速率常数。     

Am(Ⅲ)在铁氧化物上的吸附行为

为了解放射性核素在可能作为高放废物固化体包装容器材料腐蚀产物上的吸附行为,以我国高放废物处置库预选场址--甘肃北山地区深部地下水为介质,研究了包装容器材料的主要组分铁的腐蚀产物Fe2O3, Fe3O4对Am(Ⅲ)的吸附,讨论了pH值、总CO2-3, SO2-4、腐殖酸、Am(Ⅲ)浓度等对吸附的影响,并就可能的吸附机理进行了探讨.实验结果表明,Am(Ⅲ)在铁氧化物上的吸附分配比随水相pH值增大而增大;地下水的化学组分是影响Am(Ⅲ)存在形态和吸附的关键,Am(Ⅲ)在Fe2O3和Fe3O4上的吸附机理为界面配合,可用Freundlich吸附等温式描述.     

离子液体体系的萃取行为及其在乏燃料后处理中的应用前景

离子液体由于其特有的性质,在乏燃料后处理中的应用已受到广泛关注。本文综述了不同种类离子液体中多种萃取剂对乏燃料所含若干锕系元素及放射性裂片元素的萃取,重点分析了不同萃取体系的萃取效率、萃取选择性、萃取机理和反萃等关键问题。综合目前的研究成果,可发现:离子液体-萃取剂体系由于其独特的萃取机理,通常比传统萃取体系具有更高的萃取效率;一些萃取体系具有高选择性使其在乏燃料后处理中有很好的应用前景。在简要介绍阳离子交换机理、阴离子交换机理和中性复合物机理三种离子液体体系萃取机理的同时,重点总结了萃取中三相问题和协同萃取效应。本文还总结了液-液反萃、超临界CO2反萃和电化学反萃三种常见的反萃方法,讨论了各自的优缺点。本文最后对离子液体在乏燃料中的应用进行了总结与前景展望。     

溶剂萃取法在乏燃料后处理中应用研究概况

文章介绍溶剂萃取法在乏燃料后处理研究中的发展概况,主要阐明了乏燃料后处理工艺流程的研究发展过程、萃取剂的研究现状以及各种萃取剂的性能特征,并就酰胺类萃取剂的应用前景进行了概要评述。     

用于90Sr萃取分离的冠醚试剂的辐射化学研究进展

长寿命裂变产物的分离不仅可有效提高乏燃料后处理与高放废物处置安全性,而且可降低高放废物对人类和环境的危害.鉴于用于分离长寿命裂变产物的萃取剂在分离过程中通常处在较高的辐射场中,因此为了设计合适的萃取分离体系,有必要开展对裂变产物具有良好萃取性能的萃取剂的辐射稳定性的研究.本文通过对比辐照前后萃取剂的化学结构和萃取性能的...     

我国高放废物地质处置战略研究

随着我国核能事业的飞速发展,高放废物的处理和处置,将成为一个重大的安全和环保问题。这体现在最终如何安全处置核电厂乏燃料后处理产生的高放废物、核武器研制和生产过程中已产生的高放废物,以及我国存在的某些可能不宜后处理的乏燃料。     

"锕系元素及其裂变产物的超分子识别与组装化学"专栏北大核心CSCD

作为乏燃料的主要组成成分,锕系元素及其裂变产物是乏燃料后处理与高放废物处理处置过程的重要对象.在这一过程中,如何实现锕系元素及其裂变产物的高效识别、选择性分离和稳定固化是核能长期安全、高效、可持续发展需要解决的关键问题.     

从混合裂变产物中放化分离132I

依据混合裂变产物中碘及其母体碲的同位素的半衰期设计分离¹³²I的流程。该流程的主要步骤为浓HBr蒸发和CCl₄萃取。实验研究了浓HBr蒸发对碘的去污效果；在硝酸介质中，用含I₂的CCl₄作为萃取剂，研究了HNO₃浓度、水相中KI含量和有机相CCl₄中I₂含量对¹³²I萃取率的影响，测定了含SO₂水溶液对¹³²I的反萃率。用设计的推荐流程获得了放化纯的¹³²I，其中含有的¹³¹I的活度为¹³²I的1.3％，分离流程全程对¹³²I的化学回收率约为60％，流程对主要γ核素的去污因子大于10³。     

一种干式外圆机械切割装置的研制

在乏燃料后处理中，需要回取已封装在乏燃料贮存容器中的乏燃料。根据热室使用环境及乏燃料贮存容器的特点，从耐辐射设计、乏燃料贮存容器固定、切割进给、切割刀具及刀具更换、放射性废物最少化等方面进行设计响应，研制了一种在热室内开启乏燃料贮存容器的干式外圆机械切割装置。功能性试验验证了该装置满足设计和使用要求。      

高温气冷堆乏燃料后处理首端技术研究进展

高温气冷堆乏燃料采用后处理路线能充分利用核资源并减少需要最终地质处置的核废物量,有利于核能的可持续发展。传统的LWR乏燃料后处理首端过程不能用于处理高温气冷堆的乏燃料。高温气冷堆乏燃料元件及包覆层颗粒的破碎是首端处理技术的难点。破碎乏燃料元件及去除石墨的方法主要有机械碾碎法、燃烧法、脉冲电流法等;破碎及去除碳化硅的方法有传统机械碾碎法,以及正在发展中的熔融法、气流喷射粉碎法等,其中,气流喷射粉碎法具有较好的发展前景。目前,尚无一种理想的技术来解决高温气冷堆乏燃料后处理中的首端过程问题,需进一步开展高温气冷堆乏燃料后处理技术的研究。     

Management of radioactive wastes from the operation of nuclear power plants

A prerequisite for the acceptance of the nuclear energy system is the effective management of the rad-wastes. Among the wastes to be considered, there are the wastes from the operation and decommissioning of nuclear power plants, as well as those from the nuclear fuel cycle. For the management of operating wastes, processes and facilities optimized in the course of several decades, are available, with which the raw solid and liquid wastes can be reduced in volume and turned into products which are physically and chemically stable and thus suitable for final disposal. The management of spent fuel can be done either by direct final disposal or reprocessing. The required interim storage facilities are ready for operation. The methods and a facility for packaging spent fuel for direct final disposal are in an advanced stage of development and construction. If fuel assemblies are to be reprocessed abroad, the wastes generated from the process must be taken back. Decommissioning wastes have technical properties which correspond essentially to the various groups of operating wastes and can thus be processed with similar methods; however since large quantities of them are generated in relatively short times, they present particular logistic problems. All waste types end up in final disposal sites to be built under the responsibility of the federal government. A final disposal site for low level wastes is in operation. In addition, two final disposal projects for accommodating higher level wastes including spent fuel for direct disposal and vitrified wastes from reprocessing, are being pursued.     

Comprehensive evaluation of the state of spent VVR fuel assemblies from first-generation nuclear submarines

New, unconventional methods developed at the Russian Science Center Kurchatov Institute for evaluating the state of spent VVR fuel assemblies from submarines are described. They were used in combination with other methods during inspections of spent fuel assemblies held in TK-6 and -11 containers at the point of storage of spent nuclear fuel and radioactive wastes in the town of Gremikha in Murmansk Oblast.     

乏燃料密封容器开盖及其内容物回取技术研究

对于采用干湿法贮存的乏燃料而言,其后处理时面临的最大问题是如何安全高效地将乏燃料等内容物从封焊的密封容器中取出。针对这一问题,基于乏燃料密封容器及其内容物的结构特点,开展了乏燃料密封容器开盖及内容物回取技术研究,综合考虑切割热室使用环境、内容物回取后的收集和转移以及产生废物的收集处理等因素,制定了合理可行的开盖及回取工艺,研制了用于开盖和筒体分段切割的解体装置以及回取和吊装工具,并通过试验验证了工艺的可行性以及研制的工装具的可用性。      

Radiation Characteristics of Spent Nuclear Fuel and Wastes in Nonaqueous Reprocessing Methods

The effect of the technology used to separate elements with nonaqueous radiochemical reprocessing of fuel on the radiation characteristics of the fuel and the wastes is examined. The study was conducted for the intrinsically safe BREST-1200 fast reactor. The conditions for achieving radiation equivalence between the raw material used for reactor operation and the wastes of the nuclear fuel cycle are indicated. The contribution of various nuclides to the potential biological hazard of wastes is examined for various holding periods. When the requirements imposed on the radiochemical technology are satisfied, the effect of the technological differences in the separation of individual elements from the spent fuel essentially did not greatly affect the results. 4 figures, 3 tables, 6 references.Scientific-Research and Design Office of Power Engineering     

Characteristics of Several Equilibrium Fuel Cycles of PWR

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium.     

陶瓷电熔炉在动力堆高放废液玻璃固化中适用性分析

随着我国核电事业发展和核燃料循环体系日益完善,玻璃固化动力堆高放废液的需求已提上日程。为探讨陶瓷电熔炉技术在我国后续动力堆高放废液玻璃固化项目中的适用性,本文从源项、熔炉技术特点和熔炉更换解体三个方面进行了分析,认为陶瓷电熔炉技术可以用于玻璃固化动力堆产生的高放废液。     

高放废物处理、处置的国际现状分析

本文评述了高放废物处理、处置的国际现状,包括:乏燃料的后处理、贮存和直接处置;高放废液的固化方法和高放废物的处置方法。     

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.     

海洋核动力平台乏燃料贮存格架设计及安全分析研究

针对海洋核动力平台乏燃料组件海上长期贮存所面临的安全保证问题,通过改进燃料组件与贮存小室之间固定形式、优化贮存小室与贮存格架本体之间连接形式以及增加贮存格架与乏燃料水池池壁之间的缓冲结构,设计了一种满足设计基准以及适应海洋环境的乏燃料贮存格架,并采用蒙特卡罗程序MCNP-5、计算流体力学软件Fluent 14.0、有限元分析软件ANSYS 17.0对该贮存格架进行临界、热工、结构仿真计算。结果表明,该贮存格架设计合理、安全性高,可为海上浮动式核电站乏燃料贮存提供解决方案。      

重水堆密集化乏燃料干式贮存设施热工安全研究

为解决秦山第三核电有限公司（简称:秦三厂）计划延寿导致乏燃料增加、已有乏燃料干式贮存模块容量不足的问题,在原有的1~6号（QM-400）乏贮模块基础上,研发了密集化乏燃料干式贮存设施（M1型乏贮模块）。与QM-400乏贮模块相比,M1型乏贮模块贮存容量更大,能量密度更高。为论证M1型乏贮模块的热工安全性,采用RELAP5/MOD3程序,根据保守的初始假设条件建立其热工分析模型,对极端气候条件下模块正常运行和事故工况下各区域温度进行了计算,同时采用了三维流体计算流体力学（CFD）程序对RELAP5程序计算结果进行了验证,综合RELAP5程序和CFD程序的计算结果,论证了M1型乏贮模块的热工安全。