核电厂乏燃料组件厂内转运关键设备研制

乏燃料组件厂内转运是解决核电厂燃料水池贮存空间不足问题的方法之一。本文分析了乏燃料组件厂内转运的设计准则、安全风险,介绍了用于运输容器内破损组件检测和运输容器内组件冷却用设备的工作原理及其应用情况。应用结果表明:破损检测设备可以快速有效地检测乏燃料运输容器内是否存在破损组件;乏燃料组件冷却设备可以较为安全地冷却装有乏燃料组件的运输容器。      

浅析乏燃料转运容器制造过程中的设计改进

乏燃料转运容器是核电站乏燃料离堆干法贮存工程中的主要设备之一,用于将装载有乏燃料组件的密封贮存罐转运至混凝土贮存系统中,为整个转运操作提供辐射屏蔽和结构支撑.通过应用鱼骨图等质量工具,对转运容器的耳轴结构进行设计优化,调整耳轴加工工艺,在保证结构强度的前提下,提高了设计质量,保证了后续样机制造的顺利执行.     

乏燃料运输容器内破损组件检测方法

在乏燃料贮存和运输阶段,卸出时完好的燃料组件存在破损可能,若出现破损,裂变产物就有可能向环境释放,因此乏燃料运输容器解除密封前,需要进行运输容器内破损组件检测工作。本文提出一种破损检测方法,系统介绍了检测工艺方案、破损标志核素的确定及破损检测的可行性分析,并给出破损检测装置的详细设计。该设备已经研制成功并应用于大亚湾核电站乏燃料组件厂内转运,在国内首次检测到乏燃料运输后破损,结果表明,此检测方法可显著提高破损检测的可信度和效率,具有较高的工程应用价值。     

设备跌落分析报告

依据《放射性物质安全运输规定》（GB11806--89）、IAEA《放射性物质安全运输条例》（TS-R-1）、《ASME锅炉和压力容器规范》第Ⅲ卷第一册附录，对设备进行自由跌落分析与评定。设备跌落到地面的速度为53m／s。跌落分析采用ANSYS有限元程序LS—DYNA模块完成。LS—DYNA是以显式为主、隐式为辅的通用非线性动力分析有限元程序。跌落分析时，采用壳元模拟容器壁，其它的质量采用质量单元模拟，目标面采用三维结构实体单元模拟。     

液态金属反应堆湿式燃料贮存桶内部事件概率安全分析

液态金属冷却剂在给反应堆带来运行安全与热效率优势的同时,也给反应堆带来了复杂的换料系统,其中大型液态金属反应堆采用的湿式乏燃料贮存桶是乏燃料卸料过程的核心设备,临时装载了大量的乏燃料组件,具备一定的安全风险。本文采用概率安全分析（PSA）方法对乏燃料贮存桶进行风险评价,通过运行状态分析、始发事件分析、事故序列分析以及简单的定量化,初步获得其导致乏燃料组件发生损伤的事故序列和最小割集,识别了关键系统与设备。结果表明,相对于反应堆本身的风险,乏燃料贮存桶本身风险虽低但依然不可忽略,且风险评价结果对反应堆的运行方式以及清洗系统的可靠性较为敏感。此外还对该系统的设计改进与安全优化进行了讨论。     

乏燃料干式贮存技术比较分析

乏燃料贮存方式主要有湿式贮存和干式贮存两种。本文主要调研了国内外乏燃料干式贮存的概况,研究了具有贮存功能的混凝土筒仓式和具有贮存运输功能的金属容器式乏燃料干式贮存设施的技术特点,并对二者的功能、关键技术等方面进行了比较分析,指出了干式贮存技术存在的主要安全问题。最后,结合我国目前乏燃料的离堆贮存需求对我国未来乏燃料干式贮存工作提出了建议。     

基于多孔介质和等效热导率模型的快堆乏燃料组件传热特性数值分析

乏燃料组件在运输或转运过程中,组件会裸露在传热较差的气体介质内,需关注其散热性能。为模拟乏燃料组件的传热特性,采用多孔介质模型模拟组件的流动阻力,并利用等效热导率模型模拟组件内部的传热。由于自然对流条件下乏燃料组件内部流动符合层流假设,在多孔介质阻力模型中忽略了惯性力项的作用。将等效热导率模型的模拟结果与SNL-LMFBR实验结果进行对比,证明了该模拟方法的有效性。计算结果表明,在水平放置工况下乏燃料组件温度轴向呈对称分布,在竖直放置工况下轴向呈非对称分布,乏燃料组件的高温区域向组件上方偏移。     

乏燃料贮存格架时程分析方法

乏燃料贮存格架是用于贮存换料后的乏燃料组件的重要设备,其自由放置在核电厂乏燃料水池中,在地震载荷下的响应属于非线性响应,包含了各种复杂的运动:滑移、碰撞、扭转、倾覆等。为了准确描述上述非线性响应,本文建立了乏燃料贮存格架整池有限元模型并进行非线性时程分析,考虑了滑移、碰撞、摩擦等影响因素,同时还考虑了乏燃料贮存格架在水池中的流固耦合效应。通过时程分析得到了乏燃料贮存格架在地震载荷下的位移、载荷等计算结果。该方法可用于乏燃料贮存格架的抗震分析。     

压水堆乏燃料干式贮存的安全考量

通过对压水堆乏燃料干式贮存的准备、运输和贮存三个工艺区域的设备设施、操作过程和环境及其潜在的危险因素的研究, 重点针对作用于干式贮存设施的临界安全、放射性物质的包容、衰变热的移除和辐射防护的影响因素开展分析, 提出了针对性的防控对策;同时总结了秦山第三核电厂重水堆乏燃料干式贮存实施过程安全操作和管理的经验。研究成果可为解决压水堆乏燃料干式贮存设施的规划、设计、建设和运行过程有关安全问题提供思路。     

美批准扩大欧安诺干式贮存系统使用范围

正【欧安诺美国公司网站2019年4月9日报道】美国核管会(NRC)近日批准修订欧安诺公司(Orano) NUHOMS干式乏燃料贮存系统许可证,准许使用该系统贮存更多类型乏燃料,包括拥有更高衰变热的乏燃料。未来,在将乏燃料在贮存水池中存放至少2年之后,就能将乏燃料从水池转移至干式系统,从而能大幅缩短乏燃料水池冷却的时间,大幅提升乏燃料转运作业的灵活性。     

基于SCALE的RFA改进型燃料组件堆内贮存临界安全分析

RFA改进型燃料组件是西屋公司设计的能应用于大功率先进压水堆的改进型燃料组件。SCALE计算程序是一款在国际上得到广泛认可的综合性建模及模拟程序包,可用于核设计与核安全分析。基于SCALE计算程序,针对大功率先进压水堆的乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,分析乏燃料水池正常贮存及事故工况下的临界安全。计算结果表明一区正常贮存工况keff值为0.901 29,组件跌落事故工况下,有效增值因子为0.907 93。二区正常贮存工况下,计算模型keff值为0.909 98,新燃料组件误插入事故工况keff值为0.924 07。先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。     

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。      

我国乏燃料运输现状探讨

随着我国经济的持续发展,核能作为安全、清洁能源在我国能源战略中地位日益突出。在保证安全的前提下,我国核电机组按照国家规划合理增加,乏燃料的产量也将逐步增加。根据我国核电站乏燃料贮存及外运规则,以及我国核电站主要位于东部沿海,而乏燃料后处理厂处在西北腹地这一国情,必将面临乏燃料的大量、长距离及安全运输的问题。乏燃料运输作为联接核电站与后处理厂或最终处置场的纽带,在维持核燃料循环体系的正常运行上发挥至关重要的作用。对国内外乏燃料运输涉及的运输方式、运输容器、运输安全监管及事故应急体系等问题进行了分析和讨论,对我国乏燃料运输中存在问题的解决提出了建议。     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

Effect of irradiation on the impact and seismic response of a spent fuel storage and transport cask

The spent fuel storage and transport cask must withstand various accident conditions such as fire, free drop and puncture in accordance with the requirement of the IAEA and domestic regulations. The spent fuel storage and transport cask should maintain the structural safety not to release radioactive material in any condition. And also the effects of the irradiation should be considered because the spent fuels stored in the cask for a long time and be possible to change the mechanical properties of the cask.In this study, the changed mechanical properties of the cask after irradiation for the 30 years storage periods are assumed and applied to the impact analysis using ABAQUS/Explicit code and seismic analysis using ANSYS code. The stress intensity on each part of the cask is calculated and the effects of irradiation are studied and structural integrity of the package is evaluated.     

乏燃料组件提升翻转机结构设计及动力学分析

为实现乏燃料组件提升和翻转动作同步进行，本文设计了基于双导轨提升翻转同步机构的提升翻转机，并在结构中设置安全防坠单元，保证紧急情况下乏燃料组件的安全性。为进一步保证乏燃料组件在提升、下降正常工况和钢丝绳断裂异常工况下的结构安全问题，以双导轨提升翻转同步机构为研究对象进行了动力学仿真分析，使用ADAMS软件分析其多体机械系统关键部件的运动规律与运行中复杂的受力特性，获得了不同工况下设备运行全程的速度及加速度特性曲线。结果表明：双导轨提升翻转同步机构工作时乏燃料组件所受的冲击力小于10 kN，翻转提升过程中翻转料仓的速度波动小于10%，仿真结果验证了系统设计的合理性和可行性。     

Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite-moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shutdown at the end of 2004, while Unit 2 is foreseen to be shutdown at the end of 2009. At the Ignalina NPP Unit 1 remains approximately 1000 spent fuel assemblies with low burn-up depth. A special set of equipment was developed to reuse these assemblies in the reactor of Unit 2. One of most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined by using scaled experiments.The objective of this article is the estimation whether the proposed design of shock absorber fulfils the function of the absorber and the optimization of its geometrical parameters using the results of the performed investigations. Static analytical and experimental investigations are presented in the article. The finite element code BRIGADE/Plus was used for the analytical analysis. The calculation model was verified by comparing the experimental investigation and simulation results for further employment of this finite element model in the development of an optimum design of shock absorber. Static simulation was used to perform primary optimization of design and dimension of the shock absorber.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.