Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.

A Novel Computerized Water Level Control System of PWR Steam Generator of Nuclear Power Plant

This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).

Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant(BNPP)

After the Fukushima disaster, interest in the evaluation of severe accidents in nuclear power plants and off-site consequences has significantly increased. Because experimental studies are difficult to conduct, computational methods play a substantial role in accident analysis. In this study, a severe accident in the Bushehr pressurized water reactor power plant caused by a station blackout with a total loss of alternating current power supply has been evaluated. This analysis presents the in-core damage of fuel rods and the release of fission products as well as the thermal hydraulic response of the station components during the loss of active emergency cooling systems. In this manner, a perfect model of the Bushehr nuclear power plant using the MELCOR code is prepared. The accident progression is simulated, and the thermal responses of the fuels and hydraulic components are presented. It is shown that, without operator intervention, steam generators will become dry in approximately 3000 s, and the heat sink of the reactor will be lost. The simulation results show that at approximately 8600 s, the upper parts of the core start melting. This model calculates the shortest available time for accident prevention and proves that the time available is sufficient for operator manual action to prevent a nuclear disaster.

Project Construction and Important Technical Innovation for Qinshan Phase Ⅲ (PHWR) Nuclear Power Plant

Qinshan Phase Ⅲ(PHWR)Nuclear Power Plant,the first commercial heavy water reactor nuclear power plant in China,was the biggest trade project performed between the governments of China and Canada.As the owner,the Third Qinshan Nuclear Power Company(TQNPC)persisted in independent innovation management during the project construction,commissioning and self-dependent operation,efficiently realizing the three controls of the project,i.e.quality control,schedule control and investment control,and persisted in technical improvement on the basis of digestion and absorption of CANDU-6 technology to improve the unit safety and reliability.The project construction practice has helped China's nuclear power project management to becomeprogrammed,computerized,standardized and internationalized management from the existing basis.After completion of the project,with unit safe and steady operation as the prerequisite,TQNPC performed several technical modifications and innovations to continuously improve the unit performance.In the area of staff development,TQNPC paid much attention to cultivation of corporate culture,strengthed staff training and built up a good circulating mechanism with staff training and project construction promoting each other.Further to "Zero Breakthrough" and a new step forward of locolization successfully realized in Qinshan Nuclear Power Plant and Nuclear Power Qinshan Joint Venture Company,the improvement and developemnt of nuclear power project management level in Qinshan Phase Ⅲ(PHWR)Nuclear Power Plant provided reference for promotion of nuclear power development in China and standardized management of introducing large imported project.

Wide Range Neutron Monitoring(WRNM)System in Boiling Water Reactors(A Short Communication&Memorandum)

The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope was expanded to include the startup monitoring function with the aim of replacing both the source and IRMs(intermediate range monitors)in BWRs(boiling water reactors).The WRNMs,consisting of a newly designed fixed incore regenerative sensor and new electronics,which include both counting and MSV(mean square voltage)channels,have been tested in several reactors and its capabilities have been confirmed.The channel will cover the neutron flux range from 103 nv to 1.5×103 nv;it has greater than 1 decade overlap between the counting and MSV channels.Because of the regenerative fissile coating the sensor,even though fixed incore,has a life of approximately 6.0 full power years in a 51 kW/L BWR and similar situation has been proposed for newly designed small modular reactor such as BWRX-300 of General Electric Hitachi reactor.

核电厂高加水位异常分析及应对措施

秦山地区近年来发生多起高加疏水故障引起的降功率或功率波动事件,高压加热器的正常、稳定运行对机组的安全和效率有重要影响。本文对秦山第二核电厂60万千瓦级压水堆机组高压加热器近年来运行中的水位异常现象进行分析总结,提出相应的应对措施,以减少高加水位异常事件的发生,并在瞬态发生时以便运行人员能够及时、准确干预,防止事故扩大化。

Aging and Life Management System of Reactor Pressure Vessel

Reactor pressure vessel (RPV), the only key component that can not be replaced in nuclear power plants (NPPs), is the main barrier against the radioactive leakage. The lifetime of NPPs is dependent heavily on the life of RPV, and thus, the aging and life research on a RPV is a key factor in determining the life extension of NPPs. The purpose of this paper is to introduce an aging and life management system for an operating RPV which can be used as a reference of the lifetime extension. In order to realize the objective, an aging and life management system was developed. It is an comprehensive knowledge management system that integrates decentralized information and serves as a valuable data center. Based on the storage and management of RPV state information and operation data, this system provides real-time monitoring of important operating parameters, evaluation of irradiation embrittlement, and RPV aging assessment. Therefore, it is anticipated that the developed system can be used as an efficient tool for aging and life estimation of RPV.

热预应变对反应堆压力容器材料断裂行为影响研究

压水堆核电站反应堆压力容器(RPV)辐照脆化问题是制约其长期安全服役的主要因素,现有的美国ASME和法国RCC-M规范尚未充分考虑RPV用钢(铁素体材料)的热预应力(WPS)对断裂评价的有益影响。针对某RPV材料(16 MND 5),采用标准CT试样进行室温加载(L)、保持载荷降低测试温度(C),最后加载直至断裂(F)的测试方案(LCF的测试过程)。试验结果表明,在LCF的最后低温断裂阶段,RPV材料实际断裂韧度为基于RCC-M规范预测结果的两倍左右,也明显高于主曲线预测的断裂失效概率为95%对应的材料断裂韧度。因此,在RPV寿期末的脆性断裂评价中,考虑WPS效应会显著提高其安全性能评估裕量。

海洋核动力装备国内外发展现况与前景展望

海洋核动力装备是解决深远海资源开发中持久动力能源供给、海洋领域“碳减排”等问题的重要支撑。我国作为核电大国、海洋大国,虽然在核工业和海洋装备产业领域具有较好的优势基础,但在民用海洋核动力装备领域尚未实现“从零到一”的突破。本文基于对国内外海洋核动力装备发展实践研究,总结了海洋核动力装备的优势特性和技术策源,分析了未来海洋核动力装备发展的应用场景和主要趋势,厘清了我国发展海洋核动力装备的战略需求与问题,并提出了相关发展建议。研究认为海洋核动力装备总体呈现由军用向民用拓展、由陆地向海洋拓展的发展趋势,技术策源以紧凑型和一体化压水堆为主,装备类型近期将聚焦于海上浮动核电站和核动力破冰船。研究建议,通过顶层规划明确我国海洋核动力装备发展的重点应用场景,通过建立示范工程形成与发展需求相匹配的法规标准和监管制度等措施,突破海洋堆系统建造和核动力平台总装建造等方面的关键技术,推动海洋核动力装备高质量发展。

热电联产压水堆核电机组的控制方案研究

核能热电联产是实现“双碳”目标的重要方法。核能热电联产改造后,供需侧能量平衡关系发生了变化,需要针对热电联产改造后压水堆核电机组开展堆机协调控制方案的研究。首先,以大亚湾900 MW核电机组为研究对象,基于最佳估算程序RELAP5和仿真平台3KeyMaster建立了热电联产模型。热电联产形式包括主蒸汽集管抽汽、中间抽汽和背压机排汽。接着,以主蒸汽集管抽汽为例,通过典型扰动的动态特性分析并验证了所建立的模型。然后,提出了将总蒸汽流量折算为等效负荷的方法,对反应堆功率控制中冷却剂平均温度控制系统进行改造。最后,对设计的控制系统改造方案进行了验证。验证结果表明,该方案能够满足反应堆核功率、电功率和供能热负荷协调控制的要求。所设计的控制改造方案为热电联产压水堆核电机组的反应堆功率控制提供了参考。

压水堆燃料棒UO_(2)燃料芯块与锆合金包壳化学相互作用层研究

反应堆运行期间,锆合金包壳与燃料接触后不断氧化,与燃料结合形成牢固的化学相互作用层,影响燃料间隙热导、包壳力学性能和燃料包壳机械相互作用。本文以压水堆核电站燃耗45 GWd·t U^(-1)完整燃料棒为研究对象,利用金相显微镜(Metallographic Microscope)、扫描电子显微镜及能谱分析(Scanning Electron Microscopy-Energy Dispersive Spectroscopy,SEM-EDS)和热室内拉曼光谱(Raman Spectroscopy)方法对其化学相互作用层形貌及结构进行分析,国内首次获得堆内辐照后包壳和芯块化学相互作用层相关分析数据。结果表明:运行至45 GWd·t U^(-1)燃耗后,燃料芯块与包壳间隙形成14~19μm的化学相互作用层,不同位置机械接触的时间顺序差异,导致作用层的不连续形成与长大。SEM-EDS结果表明,相互作用层呈“蠕虫”状形貌,且由U、Zr、O三元素构成形成混合相(U,Zr)O_(x)化合物,并且发现化学相互作用层由化学黏附和机械作用共同作用的结果。拉曼光谱显示,化学相互作用层主要由四方相氧化锆(t-ZrO_(2))和单斜相(m-ZrO_(2))相组成。

核电厂地下综合管廊火灾隐患及自动灭火系统研究

本文重点分析了核电厂综合管廊火灾的起因、危害性及火灾特点,比较分析了在综合管廊设置各种灭火方式的优劣性,结合高压细水雾实体火灾灭火试验,在国内某核电厂地下综合管廊设计使用高压细水雾灭火系统。此项研究成果和实践对核电厂安全生产具有一定的参考价值。

核电高压给水加热器启动过程建模与仿真

为了研究金属材料的热惯性和温度变化速率对高压给水加热器启动过程的动态特性及能耗特性的影响,本文建立了基于两相非平衡模型的核电站高压给水加热器仿真模型。针对高压给水加热器的启动过程开展仿真研究,以抽汽阀门开度和给水焓值为入口扰动边界条件,分析了最大温度变化速率和正常温度变化速率下的启动过程动态特性。研究发现:高压给水加热器仿真模型完成启动后,壳侧压强、水位和给水出口温度等主要参数与设计值误差在1%以内;为了满足温度变化速率的限制,抽汽阀门开度变化率最快不能超过75.34%/h,正常变工况运行不得超过37.24%/h,对高压给水加热器的启动过程有重要指导意义。该仿真模型可以为高压给水加热器启动过程的调试和运行优化提供参考。

海洋条件下小型堆稳压器液位智能预测研究

浮动核电站的稳压器等设备受海洋条件影响发生空间位置变化并产生液面晃荡,导致现有压差测量方法不能反映容器内真实液位且易造成信号误触发。为研究海洋条件下稳压器的压差波动规律,优化液位测量方案,本文开展摇摆条件下稳压器液位测量实验;基于实验数据使用麻雀搜索算法优化极限学习机模型,构造了一种液位信号与摇摆角度、摇摆角速度、重力方向加速度、以及空间测点位置等海洋条件之间的回归预测模型。结果表明:本文所提出的麻雀搜索算法优化极限学习机预测模型具有速度快和精度高等优点,预测结果与实验数据在多种典型场景下均吻合良好,可为浮动核电站液位遥测及人工智能技术应用提供参考。

压水堆核电厂燃料包壳破损判断准则研究

燃料包壳为压水堆核电厂放射性包容的第一道屏障。因燃料包壳破损的发生不可避免且其对核安全的影响显著,燃料包壳破损诊断是压水堆核电厂必备的技术。分析了国内外压水堆核电厂燃料包壳破损诊断方法以及存在的问题,提出了根据一回路冷却剂中放化指标判断压水堆核电厂燃料包壳是否发生破损的系列判断准则,并对可能影响诊断结果的因素进行了探讨。基于国内在役核电厂实际的运行数据对判断准则进行了测试,测试结果表明,提出的燃料包壳破损判断准则可准确地诊断燃料包壳破损的发生,且有更广泛的适用性。

核电厂T3试验电路电阻的取值研究

为避免电流型闭锁T3试验偏差引入影响核电厂安全、稳定运行的风险,对电流型闭锁T3试验原理、试验电路及试验电路中应用的器件进行了研究。通过计算电流型闭锁T3试验关键步骤的电流,并建立试验开关触点所在支路总阻值R与电流表示数比值K的关系曲线,找到了一种优化原设计中电流型闭锁T3试验电路电阻取值的方法。在一定范围内,此方法可消除电流型闭锁T3试验电路中选择开关触点接触电阻的异常增大带来的试验偏差,同时降低了电流型闭锁T3试验电路对选择开关性能的要求。该研究对在役核电厂的升级改造具有较大的参考价值,同时为核安全级控制系统定期试验的设计及优化提供了新的方向。

核电厂稳压器液位控制优化分析研究

某压水堆核电厂调试期间,在执行稳压器液位控制系统定值扰动试验过程中,发现控制系统存在积分饱和现象,导致超调较大、稳定时间较长的问题。对稳压器液位原控制策略进行了剖析,结合定值扰动试验响应结果和控制器运算规律,分析了控制系统产生积分饱和的根本原因,并设计了积分分离、调节器限制替代这两种抗积分饱和优化方案。在核电厂中的应用结果表明,优化方案解决了稳压器液位控制系统中的设计缺陷,提高了核电厂瞬态响应能力和运行水平。该研究成果显著改善了稳压器液位控制系统的动态响应特性,对于核电厂的控制优化设计具有重要的参考意义,也为解决自动控制系统积分饱和问题提供了借鉴。

核电设备氢爆可用性验证试验装置的设计

针对核电安全设备有关的氢爆可用性验证一直存在无法真实模拟的难点,分析了影响核电设备氢爆可用性验证试验的因素。针对这些影响因素进行验证试验。在获得试验数据后,针对这些影响因素制定了相应的解决措施,并设计了核电设备氢爆可用性验证试验装置。该装置可模拟核反应堆失效时设备所处的极端环境。通过氢气燃爆同时实现核电厂严重事故环境条件下温度与压力曲线的包络,比用解耦的方式分别对温度、压力进行模拟更贴近实际工况,可准确控制氢气燃爆产生的温度与压力,从而解决氢爆后温度迅速跌降的问题。该设计可实现温度曲线的包络,确保试验舱内氢气燃爆后温度均匀。

压水堆核电站一回路冷却剂系统可靠性分析

压水堆核电站一回路冷却剂系统运行中存在主泵、稳压器等设备随时间失效的情况,继而会带来系统稳定性问题.根据一回路冷却剂运行特点,针对一回路主泵、稳压器、蒸汽发生器和阀门等设备,开展压水堆核电站一回路冷却剂系统可靠性研究,利用适合处理时间相关性的GO-FLOW法建立冷却剂系统的模型,对压水堆核电站一回路冷却剂系统进行可靠性分析.分析结果对于压水堆核电站一回路冷却剂系统的可靠运行提供了研究基础,对核电站安全运行具有一定的指导意义.

压力容器水位测点信号扰动问题研究

反应堆压力容器水位测量结果反映堆芯装水量的变化,用于监测堆芯是否裸露失水,为事故后工况下的一回路热工水力状态提供重要信息。某核电厂热态功能试验期间,一回路首次充水过程及半环路液位运行工况下,出现堆芯上部水位低报警信号频繁跳变问题,干扰机组操纵员对于一回路水位状态的判断。通过对压力容器水位测量原理进行深入研究,并根据水位探测器组件所处热工环境,对水位信号扰动问题作传热分析,确定该问题的根本原因。制定了优化水位测量系统加热电流函数的试验方案。结合实测数据进行分析,论证了优化加热电流方法的可行性,并给出解决方案。鉴于水位测量技术成熟,在仪控领域应用较广,该研究成果可为液位监测系统异常问题的处理提供借鉴。