核电厂继电器加速老化分析及管理研究

通过对大亚湾和岭澳核电站中压配电盘的中间和出口继电器老化与失效数据进行统计,分析核电厂继电器加速老化的主要因素。继电器加速老化的影响因素为运行环境,主要是湿度、温度和振动冲击;老化机理为氧化腐蚀、机械磨损、电磨损、有机气体腐蚀、绝缘老化和电磁干扰。建议对继电器分类管理、替代使用,进行周期检测、更换或维护。     

大亚湾核电站中压配电盘中间和出口继电器加速老化原因分析和管理对策

针对大亚湾核电站中压配电盘的中间和出口继电器加速老化问题,结合采集的老化数据和国外经验反馈,详细分析了老化机理,总结了继电器加速老化的根本原因,并对目前继电器的老化管理和剩余寿命进行了评估.根据分析结果并结合核电厂的实际,提出了继电器老化管理流程和建议,为解决核电厂继电器加速老化的管理问题提供了新思路.     

热老化对核电厂一回路管道失效概率的影响

建立基于热老化的管道失效概率计算流程。在实验研究的基础上,采取美国阿贡实验室的流程预测某管道材料在280、330℃下热老化后断裂韧性随运行时间的变化。计算含单个环向内表面裂纹的管道在考虑热老化与不考虑热老化2种情况下的累积失效概率。计算结果表明,考虑热老化因素得到的失效概率高于未考虑热老化的情况。在考虑热老化的情况下,较高的温度下热老化严重且管道失效概率更高。     

大亚湾核电站继电器机架支撑改造初步力学分析

针对大亚湾核电站继电器机架橡胶减震器因老化破裂而影响机架的抗震性能、日常运行和维修等问题而提出支撑改造方案,采用谱分析法对改造前后的各种支撑结构进行初步力学分析.比较了大亚湾继电器机架改造前后支撑的抗震性能,以及大亚湾和岭澳两种继电器机架支撑结构在同一地震载荷作用下的力学响应,为大亚湾核电站继电器机架支撑的加固改造提供参考.     

阀门组件老化敏感点预警值的确定及老化缓解对策

对大亚湾核电站死管段相关的核一级电动阀阀门组件的机械老化机理进行了讨论,并针对阀门组件的失效模式,对老化敏感点进行了识别,选出最重要的老化敏感点进行分析讨论,确定其预警值,最后提出并归纳了一些适合大亚湾核电站实际情况的老化缓解对策。     

秦山核电厂堆内构件围板螺栓老化评估

介绍了秦山核电厂为评估堆内构件围板螺栓的实际老化状态,在吸收总结国际上堆内构件老化机理研究成果的基础上建立了堆内构件围板螺栓的老化机理判断准则,用其评估识别围板螺栓主要受磨损、应力松弛、辐照肿胀、辐照促进应力腐蚀开裂等老化机理的影响,并针对老化机理可能导致的缺陷类型,开发了水下超声检查技术补充常规的目视检查方法,从而制定评估老化状态的检查方案。评估结果表明,秦山核电厂堆内构件围板螺栓老化状态良好,尚未发生变形、裂纹等老化失效现象。实践证明该评估方法行之有效,可用于压水堆核电厂部件老化评估工作。     

反应堆压力容器用Ni-Cr-Mo-V钢焊缝热老化脆化行为研究

对反应堆压力容器用Ni-Cr-Mo-V钢焊缝温度监督样品的热老化脆化行为进行了研究。焊缝属于压力容器的薄弱环节，服役时间最高达120 430 h（服役温度归一化到300 ℃）。3批次的焊缝监督样品冲击实验表明，焊缝材料在热老化过程中发生了脆化。通过研究发现，金相组织和显微维氏硬度在热老化期间未发生明显的变化，表明在热老化过程中不存在硬化脆化机制。断口分析及扫描俄歇纳米探针研究表明，晶界发生了P的偏析，弱化了晶界结合力，因此，反应堆压力容器用Ni-Cr-Mo-V钢焊缝在热老化过程中发生了由杂质元素P偏析引起的非硬化脆化。     

应用于核电厂老化PSA的SSC筛选方法研究

核电厂运行许可证延续必须考虑其延寿期内的核安全问题,确保核电机组在延期运行期间的核安全水平不低于原设计寿期内的核安全水平。可应用PSA技术对许可证延续期间的核电厂建立老化PSA模型,从而评估SSC老化对核电厂整体安全的影响,验证其仍可满足原设计标准。基于此提出了应用于核电厂老化PSA的SSC筛选分析方法,通过考虑趋势分析,老化失效模式与影响分析,风险重要度分析,在三种分析方法基础上建立核电厂SSC筛选的决策矩阵,为选择易老化且安全重要的部件建立了可行的方法。该项工作也为核电厂在许可证延续阶段的风险指引型管理奠定技术基础。     

浸渍活性炭性能和微观结构老化效应研究

为了研究老化引起的核级浸渍活性炭性能、结构上的变化,对现场应用老化及同批次自然老化后的浸渍活性炭,开展了吸附效率、关键物理性能、微观结构、热稳定性等变化特征的实验分析。研究发现,自然老化54个月的浸渍活性炭各性能指标变化不显著,但对于现场老化后吸附放射性甲基碘效率降至不足60%的浸渍活性炭,其CCl₄吸附率由初始的近60%显著下降至15%,pH值从9.7下降至7.2,碘吸附值下降至新浸渍活性炭的3/4,与除碘效率的下降趋势一致;SEM、N₂吸附测试表明,现场老化后的样品微观孔结构变化显著,孔道堵塞、磨损严重,比表面积减小,孔径增大;TG(失重实验)结果表明,现场老化后的浸渍活性炭在200~500 ℃失重约10%。这些理化性能及自身结构的变化是造成浸渍活性炭老化及除碘性能下降的重要原因。     

基于性能退化的热老化可靠性剩余寿命预测方法

以核电厂主管道为研究对象,运用性能退化可靠性理论和贝叶斯更新方法,对主管道因热老化导致失效的剩余寿命进行研究。首先通过加速热老化实验获得的数据,分析主管道奥氏体不锈钢材料的冲击性能退化过程,然后运用贝叶斯更新方法得到基于冲击性能衰退信息的主管道热老化剩余寿命分布,进行可靠性剩余寿命预测;最后通过实例证明该方法的有效性和可操作性。     

如何发现和解决大亚湾和岭澳核电厂存在的加速老化现象

从振动、腐蚀和有机材料材质三大方面对国内外核电厂设备加速老化现象进行了阐述。根据大亚湾和岭澳核电厂的特点,提出了目前核电厂常规岛可能存在的加速老化问题和主要影响因素,并对如何发现和准确评估已存在的老化现象,以及如何分析和研究加速老化现象和具体的解决措施进行了详细介绍。     

Understanding aging in containment cooling systems

A study has been performed to assess the effects of aging in nuclear power plant containment cooling systems. Failure records from national databases as well as plant-specific data were reviewed and analyzed to identify aging characteristics for this system. The predominant aging mechanisms were determined, along with the most frequently failed components and their associated failure modes. This paper discusses the aging mechanisms present in the containment spray system and the containment fan cooler system, which are two systems used to provide the containment cooling function. The failure modes, along with the relative frequency of each, are also discussed.     

BWR control rod drive system aging

This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of “managing” this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each U.S. BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year.Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components affected by the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms.     

Fault tree analysis of CANDU shut-down system

The CANDU shut-down system comprises electro-mechanical shutdown rods and liquid poison injection, each of which includes sensors, instrument channels and mechanical and fluidic subsystems. Published work so far has reported 10⁻⁵ for the unavailability of the reactor protection system. The basic component failure rate is assumed to be constant and to have a lognormal distribution. Mathematical models are developed to analyse the CANDU shutdown system, with aging of basic components. The Weibull distribution is used because it has a lower standard deviation. An unavailability of 10⁻⁴ is obtained. Time constraint on system operation and aging of components over a year do not significantly affect system unavailability.     

PRAAGE88: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

The effects of time dependent failure rates caused by the aging of components are becoming increasingly important in probabilistic risk assessment (PRA) and reliability analyses of nuclear power plant systems. In the NRC Nuclear Plant Aging Research (NPAR) program, the effects of aging in nuclear systems are being evaluated through the use of time varying failure rates that are determined as a function of the age of the system. These analyses involve complex systems and include various sensitivity studies; thus, the PRAAGE88 computer code was developed to facilitate these calculations. PRAAGE88 is an IBM PC based code that computes system unavailability, component unavailability, and various importance measures for use in evaluating the effect of aging on reactor systems. This paper describes the methodology utilized in the code, its capabilities and areas of application.     

Failure initiation and propagation of Li4SiO4 pebbles in fusion blankets

Lithium orthosilicate (Li₄SiO₄) pebbles are considered to be a candidate as solid tritium breeder in the helium cooled pebble bed (HCPB) blanket. These ceramic pebbles might be crushed during thermomechanical loading in the blanket. In this work, the failure initiation and propagation of pebbles in pebble beds is investigated using the discrete element method (DEM). Pebbles are simplified as mono-sized elastic spheres. Every pebble has a contact strength in terms of critical strain energy, which is derived from a validated strength model and crush test data for pebbles from a specific batch of Li₄SiO₄ pebbles. Pebble beds are compressed uniaxially and triaxially in DEM simulations. When the strain energy absorbed by a pebble exceeds its critical energy it fails. The failure initiation is defined as a given small fraction of pebbles crushed. It is found that the load level for failure initiation can be very low. For example, if failure initiation is defined as soon as 0.02% of the pebbles have been crushed, the pressure required for uniaxial loading is about 2.5 MPa. Therefore, it is essential to study the influence of failure propagation on the macroscopic response of pebble beds. Thus a reduction ratio defined as the size ratio of a pebble before and after its failure is introduced. The macroscopic stress–strain relation is investigated with different reduction ratios. A typical stress plateau is found for a small reduction ratio.     

Phase I aging assessment of nuclear air-treatment system high efficiency particulate air and adsorbers

A phase I aging assessment of high efficiency particulate air filters and activated carbon gas adsorption units was performed by the Pacific Northwest Laboratory as part of the US Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. Information was compiled concerning design features, failure experience, aging mechanisms, effects, and stressors, and monitoring methods. Over 1100 failures, or 12% of the filter installations, were reported as part of a US Department of energy survey. Investigators from other laboratories have suggested that aging could have contributed to over 80% of these failures. Several instances of impaired performance as the result of the premature aging of carbon were reported. Filter aging mechanisms range from those associated with particle loading to reactions that alter the properties of gaskets. Mechanisms that can lead to impaired adsorber performance include the loss of potentially available active sites as a result of the adsorption of moisture or pollutants. Stressors include heat, moisture, radiation, and airborne particles and contaminants.     

HTR-PM燃料元件生产穿衣工艺及设备研制

自主研制了穿衣工艺所用非标设备,包括穿衣系统、滚筒筛、振选台、均匀化设备、不合格颗粒回收系统等。工艺实验结果显示,穿衣颗粒成品率高且稳定,破损率满足设计要求。设备易于操作控制,完全能满足高温气冷堆示范电站燃料元件规模生产的需要。