低氧化态金属离子(LOMI)去污技术研究(最终报告)

简要介绍了自1986年以来对LOMI去污技术的研究:包括某些基础性实验研究及试剂制备;LOMI去污剂的脱膜效率和腐蚀速率试验;对PWR污染部件的去污试验以及放射性废水的处理方法等。实验证明NP/LOMI去污技术和所研究的废水处理技术应用于PWR是可行的,并且有一定的优越性。采用减压蒸馏技术制备低甲酸度LO-MI去污剂和采用二次电解法把钒的利用率提高到95％以上是独创的。     

压水堆核电站在役化学去污工艺进展

针对在役压水堆核电站主回路及核辅助系统污染的特点,本文介绍了国际上具有代表性的化学去污工艺HP/CORD-UV、AP-CITROX、CAN-DEREM、LOMI、EMMAC,对比分析了去污配方、去污工艺、材料兼容性、废物最小化及潜在风险等,并针对这些技术的发展和应用提出了建议。     

商业核电站退役前拆除前去污技术的开发

本文介绍了作为商业核电站退役的各种技术之一的商业核电站退役拆除前的去污技术的开发,考查研究了系统（如管道）的化学去污和研磨去污。以及大型设备和槽罐的凝胶喷射去污技术。在基本实验中,用冷模拟样品进行了去污试验,得到了最佳的去污条件。从已取得的结果来看,把它们用于商业核电站退役拆除前的去污有较好的前景。     

电化学去污工程应用的现场验证试验

采用电化学去污技术对放射性污染金属进行了现场验证试验.结果表明:在前期研究得出的电解液及工艺参数下,电化学去污技术与去污剂擦洗、酸液循环两种常用去污方法相比,可快速、高效地对放射性污染金属表面进行去污;电化学去污技术对于不同污染类型、不同污染程度的金属设备均能有效去污,在电解5 min内,放射性水平可降至仪器探测限以下;在电化学去污过程中,金属的污染水平随电解时间增加呈对数下降趋势,去污系数DF值成线性变化.     

面向强放热室退役的高效复合去污技术研究

强放热室作为反应堆材料辐照检验的配套设施,其辐射水平高、结构复杂、去污难度较大。针对强放热室退役不锈钢壳体去污的特殊性和复杂性,开展了高压水射流去污、可剥离膜去污和机械打磨去污3个单项去污试验和去污工艺试验研究,并创新性的提出了一种强放热室不锈钢壳体高效复合去污工艺。经工程去污实践验证,去污后热室不锈钢覆面表面污染水平均低于40 Bq/cm2,去污因子最高达110以上,达到了国内先进水平。热室高效复合去污技术的研发解决了强放热室不锈钢壳体表面去污的技术难题,降低了退役阶段工作人员的受照剂量,保护了工作人员和环境的安全,具有显著的经济、社会效益。      

反应型压制去污剂动态混合特性试验研究

压制去污技术是有效处置核事故污染的技术手段之一。去污材料与其相应的作业机械构成了去污技术的组成部分,目前去污剂喷洒作业过程中常存在去污材料混合不均匀、膜体不易剥离等问题。为此,开发了一款主动旋转混合方式的动态混合器,解决去污剂混合不均等问题。为验证动态混合器混合效果,搭建了反应型压制去污剂动态混合试验系统,开展了去污剂混合均匀性、成膜时间、可剥离性、膜体微观形态、拉伸强度等试验研究。试验结果表明,新型动态混合器性能较好,材料混合均匀,材料膜体力学性能和可剥离性均满足应用要求,研究结果可为去污喷洒车改进设计提供重要的技术支撑。     

放射性表面污染金属废物激光去污工艺研究

为实现激光去污技术在放射性表面污染金属废物清洁解控或循环再利用方面的应用,以350 W的纳秒脉冲光纤激光器为基础搭建了激光去污实验装置,针对激光功率、脉宽、频率、线间距、扫描速度等关键参数开展了一系列激光剥离去污工艺实验,根据实验结果分析得出激光去污工艺规律和不同去污深度的最佳工艺参数,并以某核电厂控制棒水池贮存搁架底板为对象开展验证试验。验证试验结果显示,采用激光去污技术,去污深度达到10 μm后,样品的β射线放射性表面污染水平已低于0.8 Bq/cm2,可达到清洁解控的表面污染水平要求。      

管道在线超声去污技术在某核电站的初步应用研究

核电站放射性系统水介质传输管道经过长期运行,部分管段会出现放射性物质沉积的现象。为了提升管道冲洗去污效果,国内部分核电站开始尝试引入在线超声去污设备,这种设备可安装在管道外侧,通过将超声振动传递至管道内,使内壁沉积物剥落,从而提升管道冲洗的去污效果。在线超声除垢设备在核电站的应用仍处于试验探索阶段,其功能有待进一步开发、完善。本文主要介绍国内某CPR1000堆型核电站开展管道在线超声去污的研究工作,该研究分别在60.33 mm、88.9 mm管径管道开展了试验,其中在88.9 mm管径管道试验中选取了松散污染沉积和固定污染沉积两种不同污染类型的管道,试验证明该技术在松散污染沉积管道有较明显的去污效果,具有推广和进一步研究的价值。     

核设施建筑物表面去污技术

国际原子能机构（UAEA）在相关标准中已明确指出：“在核设施设计阶段和运行阶段就应该考虑退役”。随着我国核电厂建设步伐的加快，对核设施退役技术的研究也日益受到重视。在核设施退役中，建筑物表面的去污是重点和难点，国外就此项技术开展了多年的研究和试验，并在实践中得到了应用，取得了令人满意的效果。本文对国外一些先进、成熟的建筑物表面去污技术和设备，就其在去污效果、去樗效率和二次废物产生量方面进行了分析和比较。这些技术可供我国核电厂及其他核设施退役时借鉴。     

Purex流程中酸度对锆钌净化的影响

在保证Purex流程中钌高净化效率的基础上,尽量提高锆的去污效率,以优化 Purex流程去污工艺.本文通过搭建台架试验平台,采用多级萃取方式,模拟Purex流程中共去污过程,开展了 1 AF酸度和 1 AS酸度对铀锆钌在流程中的分布情况及对锆钌去污效果的影响研究.研究结果表明:提高1 AF酸度不利于锆的去污,有利于钌的...     

Application of radiation chemistry research to the operation of nuclear reactors

The theme of this review is the application of radiation chemistry research to improve the operating efficiency of nuclear reactors. The intense radiation fields in reactor cores produce a hostile environment for incore materials; this report describes how recent research helped overcome the chemistry problems caused by the radiation.Examples discussed are the inhibition of graphite moderator corrosion and prevention of carbon deposition in gas-cooled reactors, suppression of radiolysis of the cooling water in concrete pressure vessels, hydrogen formation following a loss of coolant accident in a PWR and improving the stability of decontamination reagents for water reactors.     

基于等价理论的压水堆包壳材料共振计算方法研究

共振自屏效应的处理是影响压水堆组件程序反应性精度的主要因素之一,压水堆锆包壳材料同样具有共振自屏效应,忽略其影响会对反应性造成100～300 pcm(1 pcm=10^-5)的偏差。目前,主要通过提供经验上的参考稀释截面与包壳等价理论处理包壳材料的共振自屏效应,但并未对其适用性及精度进行完整的分析。因此,本文采用DRAGON程序,通过一系列压水堆算例对这2种方法进行测试,确定包壳共振自屏效应的主要影响因素以及这2种方法的适用性。结果表明,包壳材料的共振自屏效应仅仅与包壳区的原子核密度、厚度、慢化区的水铀比有关,并且参考稀释截面方法可以满足大部分典型压水堆系统的计算精度,但是对于包壳区尺寸、原子核密度、慢化区水铀比变化较大的系统计算精度较差,而包壳等价理论计算精度和普适性强,可用于不同类型压水堆系统包壳材料的共振自屏计算。     

压水堆核电厂采用富集硼的优势与可行性分析

压水堆核电厂通常采用天然硼进行反应性的化学补偿控制,研究表明采用富集硼替代天然硼,可以优化一回路水化学,降低一回路结构材料腐蚀风险和堆芯沉积风险,降低职业照射剂量。本文分析压水堆核电厂采用富集硼的优势与可行性,介绍国内外核电厂富集硼的应用情况。最后对我国华龙一号堆型和在役压水堆核电厂富集硼的应用提出建议。     

Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO₂ and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO₂ fueled PWR under both reactivity initiated accidents and loss of coolant accidents.     

压水堆核燃料循环的经济分析程序介绍

所开发的压水堆核燃料循环分析计算机经济程序包括12个子程序,代表着压水堆整个核燃料循环各种不同的工艺过程。本程序能算出压水堆核电站核燃料循环中燃料费用对发电成本的影响,给出各工艺过程对燃料成本的敏感度分析,并尽可能给出燃料循环中燃料材料及服务的价格数据.     

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.     

Method of Estimating Liquid Velocity in Hot Leg during Reflux Cooling of Natural Circulation in Pressurized Water Reactor

Natural circulation in a PWR has considerable attention since the TMI-2 accident as an alternative cooling method or recovery technique from certain kinds of accidents or transients involving a loss of pumped circulation. Among the three modes of natural circulation (i.e. single-phase, two-phase and reflux cooling), reflux cooling has not been well investigated in a PWR configuration. The present study was thus focused on reflux cooling of natural circulation and analytical method was developed to estimate the liquid velocity of the condensed liquid in a hot leg of a PWR.

The results of the present study showed that the liquid velocity and the liquid thickness are estimated as 2.7 m/s and 3.0 cm, respectively, at the hot leg inlet from the upper plenum for the typical PWR reflux condition (2% core power at 6.9 MPa). Therefore it was concluded that a flow-blockage of the steam flow from the core by the condensed liquid flow is unlikely to occur in a hot leg. The results are also useful for designing a special instrumentation for measuring the condensed liquid flow rate and the liquid thickness in an experimental test facility for reflux cooling test.     

Control Rod Worth in High Conversion PWR

In a High Conversion PWR (HCPWR), the neutron spectrum is hardened in comparison with the conventional PWR and the nuclear characteristics of the control rod is much different. We have calculated the control rod worth in a HCPWR for several materials of strong neutron absorbers and selected B₄C for the control rod material. For the B₄C control rod, we have investigated the various nuclear characteristics such as the spatial and energy dependency of neutron absorption, the dependency of the control rod worth on moderator to fuel volume ratio and Pu enrichment etc., and obtained the useful data for the nulear design of control rods in a HCPWR.     

先进压水堆核电厂应急计划区探讨

探讨了影响核电厂应急计划区大小的主要因素并阐述了作者的技术见解,提出了在先进压水堆核电厂应急计划区范围测算过程中,严重事故截断概率取10^(-8)的建议。     

Nuclear material investigations by advanced analytical techniques

Advanced analytical techniques have been used to characterize nuclear materials at the Paul Scherrer Institute during the last decade. The analysed materials ranged from reactor pressure vessel (RPV) steels, Zircaloy claddings to fuel samples. The processes studied included copper cluster build up in RPV steels, corrosion, mechanical and irradiation damage behaviour of PWR and BWR cladding materials as well as fuel defect development. The used advanced techniques included muon spin resonance spectroscopy for zirconium alloy defect characterization while fuel element materials were analysed by techniques derived from neutron and X-ray scattering and absorption spectroscopy.