压水堆燃料棒包壳管抗蚀性影响分析

压水堆燃料元件中燃料棒包壳管采用的是锆合金材料,近年来为探索性能更好的锆合金材料,其制造工艺一直不断发展,通过调整添加不同比例的杂质元素,期望得到更适合压水堆电厂燃料组件包壳的锆合金材料,作为燃料包壳材料的发展经历了Zr-1、Zr-2、Zr-3、Zr-4等阶段。目前压水堆锆合金包壳管最多采用的是Zr-4、M5和Zirlo这三种锆合金,这些锆合金从本质上讲主要是由于所含杂质元素含量不同。为了解这些锆合金包壳管中的杂质含量对燃料棒包壳管在堆内的抗蚀性影响,对燃料棒包壳管中部分杂质元素进行了分析,找出其杂质元素对抗蚀性的影响规律,为后续新型燃料棒包壳管的研制奠定了基础。     

连铸中间包清除渣壳问题的研究

介绍了伊朗Mobarakeh钢厂中间包清除渣壳问题的研究情况。中间包渣壳清除不好则会使连铸出现问题、耐火炉衬之间发生化学反应，渣从钢包进入中间包。渣／工业镁橄榄石混合物耐火材料界面，通过采用XRF、XRD、OM、TEM、SEM和EDS技术分析显示从钢包进入中间包内的渣的体积在中间包中对耐火材料损毁有直接影响。中间包里渣壳被固定进安全衬内的百分比由炉渣碱度和中间包里的渣中的Al2O3的量决定。这显示出一个直接的侵蚀机理，即发生在碱性渣和低铝含量的酸性渣里，而间接侵蚀机理发生在高铝含量的酸性渣里。     

核燃料后处理厂厂房布置现状与展望

乏燃料后处理是实现核燃料闭式循环的关键步骤,其中一个重要难点在于物料具有放射性,对乏燃料后处理厂房的布置设计就提出了更高的要求。后处理厂房布置设计主要包括总图设计和厂房设计两部分。总图设计需考虑不同厂房放射性高低的差别,不同厂房间物流、人流的通畅,公用工程使用的便利性等,同时需要尽量减少占地面积,降低成本。厂房设计方面,需要满足工艺流程需求,便于检维修,考虑空气提升等免维修流体输送手段的使用,同时,需要尽量减小厂房整体体积,降低成本。本文对国内外后处理厂房的布置情况进行了综述分析,总结提出了后处理厂房布置设计过程中需要考虑的厂房布置要求、建筑设计要求以及安全设计要求,可为后续后处理厂房的布置设计提供借鉴和参考。     

乏燃料后处理工艺中叠氮酸的产生及危害分析研究

叠氮酸(HN_3)是一种爆炸物。在乏燃料后处理工艺过程中,支持还原剂肼(N_2H_4)大量应用于消除系统中亚硝酸(HNO_2),该过程会导致体系中产生HN_3及叠氮酸盐等。通过对N_2H_4的行为进行分析,得到HN_3在乏燃料后处理各个主要工艺环节的分布;列举叠氮酸和叠氮酸盐可能造成的事故以及事故的消除措施;针对性的提出降低料液中HN_3含量的方法,从而减少对工艺过程中的危害。     

燃料包壳破损工况下移动式废液处理系统累积剂量水平评估初探

AP1000移动式废液处理系统用于处理超标化学废液或事故工况下的放射性废液,处理后的废液满足液态流出物的排放标准。采用蒙特卡罗方法估算了AP1000核电站在0.25%燃料包壳破损情况下,移动式废液处理系统连续处理放射性废液30 d后,距离屏蔽30 cm处的剂量水平,验证了现有76 mm碳钢屏蔽的有效性,为运行人员的辐射防护提供了参考依据。     

在线分析技术在乏燃料后处理厂的应用

后处理在线分析技术是对乏燃料后处理厂工艺流程中的放射性物流直接进行在线测量的一项分析技术,能实时提供测量结果以便于快速及时调整工艺参数。本文对比了国内外以及我国不同时期在线分析技术的发展动态,分析了不同在线测量装置的主要特点,提出了我国大型乏燃料后处理厂在线分析点和对应分析方法的设置规划,以及后续在线测量装置的主要研究重点和发展方向。     

The Possibility of Reprocessing of Spent Nuclear Fuel Using Supercritical Fluids

For the first time, the possibility of dissolution of spent nuclear fuel from a nuclear power plant in liquid and supercritical carbon dioxide was demonstrated. As shown by the example of spent nuclear fuel, the dissolution and the extraction of actinides and fission products by solutions of tributyl phosphate and nitric acid adducts TBP(HNO₃)_1.8 in carbon dioxide can be used as one of the stages of spent nuclear fuel reprocessing.     

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ～3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO₃ and 0.20 M N₂H₄ is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.     

核燃料后处理厂有机溶剂的净化处理

文章调研了核燃料后处理厂萃取工艺过程中有机溶剂的降解产物种类及各种降解产物对萃取工艺过程的影响,同时,为提高有机溶剂的利用率,减少废液量,采用碱酸洗涤、精馏等净化工艺处理使用过的污溶剂,净化后的溶剂再返回萃取流程中使用,文中分析对比了污溶剂的不同净化处理方法及其优缺点。     

我国核燃料后处理仪表发展技术研究

回顾了早期生产堆后处理厂中接触式测量仪表的选用、安装和维护方式;论述了动力堆后处理中试厂使用的非接触式和间接式测量仪表的原理和优势;在分析智能仪表优势的基础上,给出了Profibus PA和FF总线仪表在后处理工程的应用;最后,得出非接触式和间接式的测量方案与智能仪表的组合是未来后处理厂最佳选择的结论。     

核燃料循环后端项目PDMS三维设计管道等级命名规则研究

通过调研与分析国内化工行业标准、国内外化工项目、核电项目、核燃料循环前端项目以及国外核燃料循环后端项目等在PDMS三维设计中采用的管道等级命名规则,并结合我国核燃料循环后端项目的工程特点,确定了核燃料循环后端项目三维设计管道等级命名规则。此外,本文也进一步对命名规则中各个字母所代表的意义进行了分析研究,最终形成了核燃料循环后端项目三维设计管道等级清单,为三维设计在核燃料循环后端项目中的实施打下了坚实基础。     

核化工后处理厂系统去污研究

早期核化工后处理厂主工艺系统在运行过程中直接接触高放射性水平物相,污染水平很高。根据现有源项调查结果,我国早期核化工后处理厂中试厂放射性污染严重。本文阐述了清洗去污的必要性和作用,进一步分析利用原有设备串联关系重新设计洗涤组的合理可行性,提出废液倒出是系统去污的关键点,给出了解决这一关键点的废液倒出方案。     

Applications of Diglycolamide Based Solvent Extraction Processes in Spent Nuclear Fuel Reprocessing,Part 1: TODGA

Over the last decade there has been much interest in the applications of diglycolamide (DGA) ligands for the extraction of the trivalent lanthanide and actinide ions from PUREX high active raffinates or dissolved spent nuclear fuel. Of the DGAs, the N,N,N’,N’-tetraoctyldiglycolamide (TODGA) is the best known and most widely studied. A number of new actinide separation processes have been proposed based on extraction with TODGA. This review covers TODGA-based processes and extraction data, specifically focusing on how phase modifiers have been used to increase metal loading and thus enhance the operating process envelopes. Effects of third phase formation and the organic phase speciation are reviewed in this context. Relevant aspects of the extraction chemistry of important solvents (TODGA-modifier-diluent combinations) are described and their performances demonstrated by a consideration of the published flowsheet tests. It is seen that modifiers are successfully enabling the use of TODGA in actinide separation processes but to date the identification and testing of suitable modifiers has been rather empirical. There is a growing understanding of the fundamental chemistry occurring in the organic phase and how that affects extractant speciation and metal loading capacity but studies are still needed if TODGA-based flowsheets are to become an industrially deployable option for minor actinide (MA) recovery processes.