Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors

In the present paper, new results using in situ video are presented regarding boiling water reactor (BWR) control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High-content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B₄C-granules prevented direct steam attack of residual B₄C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.     

Visual classification of waste material for nuclear decommissioning

Redundant and nonoperational buildings at nuclear sites go through the process of ‘decommissioning’, involving decontamination of nuclear waste material and demolition of physical infrastructure. One challenging problem currently faced by the nuclear industry during this process is the segregation of redundant waste material into a choice of ‘post-processes’ based upon the nature and extent of its radioactivity that may pose a serious threat to the environment. Following an initial inspection, waste materials are subjected to treatment, disruption and consigned to various types of export containers. To date, the process of objects (waste) classification is performed manually. In order to automate this process, robotic platforms can be deployed that utilise robust and fast vision systems for visual classification of nuclear waste material. This paper proposes a novel solution incorporating a machine vision system for autonomous identification of waste material from decommissioned nuclear plants. Rotation and scale invariant moments are used to describe object shapes in the visual scene whereas a random forest learning algorithm performs object classification. Using nuclear waste simulants (from the nuclear plant decommissioning process), an exhaustive ‘proof-of-concept’ quantitative assessment of the proposed technique is performed, in order to test its applicability within the current problem domain.     

微型反应堆卸料装置设计及可靠性分析

微型反应堆（简称“微堆”）低浓化及退役都包含乏燃料卸出的操作，而保证乏燃料安全卸出的关键设备之一就是卸料装置。现有的卸料装置在操作过程中会破坏微堆堆筒体密封性，并且无法恢复，但微堆低浓化后还需利用原有堆筒体进行装料运行，所以本文在此需求的基础上设计了一套新型的卸料装置，可在不分离筒节、不破坏筒体完整性及密封性的前提下完成卸料操作。新设计的卸料装置包含卸料操作工具和辅助机械装置两部分。卸料操作工具通过小盖开口即可实现燃料组件的抓取，实施吊装。卸出的微堆乏燃料具有很高的放射性，卸料操作工具配合辅助机械装置，可实现远距离起升平移的操作，这种设计便于屏蔽，同时可有效降低工作人员所受辐射剂量。对该卸料装置进行计算和可靠性分析，结果表明其强度远大于实际使用载荷，安全可靠，能较好地满足微堆使用需求。新型微堆卸料装置具有经济性好、易制备、易操作的特点，下一步将在国内外微堆低浓化卸料或退役中推广应用。     

退役铀矿山和水冶厂尾矿、废石治理的代价-利益分析

根据辐射防护三原则，利用代价-利益分析方法，建立了退役铀矿山和水冶厂的尾矿，废石治理的目标函数，约束函数和限值方程，提供了一种最优化的分析方法。     

压力水去污技术及其在核设施退役工程中的应用

简要介绍了低压水、高压水和超高压水去污技术和经验；针对受核污染的水池和房间的不同情况，进行了高压水喷射去污的验证性试验，并得出了一些结论；介绍了核设施退役工程中高压水喷射去污的设计、实施和效果．     

反应堆压力容器退役包装屏蔽方案初探北大核心CSCD

为保证核电反应堆压力容器安全退役,本文以国内最早运行的秦山一期反应堆压力容器源项为参考,模拟设计保持压力容器完整和切割压力容器两种包装屏蔽方案,通过估算两种方案下废物体积、包装成本、运输及处置成本,对比分析发现切割压力容器方案更佳。研究成果可为核电站退役工作提供支持。     

某铀尾渣库退役治理覆盖系统设计

结构合理的覆盖系统是有效降低铀尾矿(渣)库氡析出率的重要措施。在总结国内外铀尾矿(渣)库覆盖系统研究成果及发展历程的基础上,根据辽宁某铀尾渣库现场覆土降氡试验成果,提出该尾渣库退役治理多层覆盖系统结构型式设计,并确定了覆盖材料、厚度和组成。设计的覆盖系统结构型式简单,施工方便,满足将氡析出率降至规定限值以下的要求。     

中低放固体放射性废物处置中的α废物在线检测及位置判定方法初探

根据在核废料回取分类、压缩减容的处置工艺及部分专用设备开发等方面的研究和初步实践,提出了利用现代测试技术和自动控制原理对a型废物进行在线探测、位置判定、目标成像、定性分析、比活度粗测的方法。     

放化厂退役去污

阐述了放化厂的退役去污，介绍了去污的目的、主要污染核素及其来源和污染的机理，并论述了系统去污和拆卸去污工艺的选择等。