EPR核电厂CRDM耐压壳焊缝缺陷信号分析

核电厂控制棒驱动机构（CRDM）长期处于高温、高压和高辐射环境中,其耐压壳异种金属焊缝容易出现裂纹等缺陷,是大修期间在役检查的关注重点。针对该焊缝役前超声检查中发现的疑似缺陷显示,通过补充目视、射线、超声相控阵和破坏性试验,验证了此类显示信号为不完全再结晶的奥氏体硬化晶粒造成的冶金显示,不影响耐压壳焊缝的质量,并总结了核电厂核岛设备超声疑似缺陷信号分析验证的方法。      

CRDM中部?密封环堆焊修复裂纹扩展分析

核电厂控制棒驱动机构(CRDM)耐压壳采用?密封环焊接安装在反应堆压力容器顶盖的管座上。一回路水应力腐蚀易诱发?密封环焊缝产生裂纹导致泄漏,需采用堆焊技术进行修复以保证?密封环结构完整性。基于ASME规范中的断裂力学分析方法,针对?密封环堆焊修复的设计结构,开展疲劳和应力腐蚀引起的裂纹扩展分析计算,为?密封环堆焊设计和评定提供参考。     

Ω焊缝焊接工艺参数的正交试验设计

为了确定核电厂反应堆控制棒驱动机构（CRDM）焊接机焊接工艺参数,应用正交试验设计法进行了Ω焊缝焊接工艺评定试验,用数理统计方法分析了对焊缝质量产生影响的各焊接参数的主次顺序,得到了最优生产条件。     

核电厂蒸汽发生器与主泵对接焊缝检查系统开发

AP1000核电厂蒸汽发生器出口接管与主泵泵壳对接焊缝泵壳侧为粗晶奥氏体铸造材料,由于该焊缝壁厚大、超声衰减、晶粒散射严重等导致焊缝的超声检测技术开发难度大。本研究采用特殊的设计,开发了一套从蒸汽发生器出口接管内壁实施超声检测的自动检查系统,并将该系统应用于国内某AP1000核电厂的役前检查。结果表明,该检查系统完全满足现场检查要求,检验结果与焊缝出厂检验结果具有良好的一致性。      

核电厂一回路管座焊缝超声波检测仿真研究

小尺寸支管接头（BOSS）焊缝作为核电厂一回路压力边界的薄弱环节,对其有效监控是核电厂日常和在役大修的重点和难点。采用仿真技术、工艺试验和现场应用验证等方法,设计并验证了BOSS焊缝的超声波相控阵检测工艺,解决了核电厂日常和在役大修中BOSS焊缝的监督难点。并得到类似超声波相控阵检测工艺的设计和验证方法。      

核电厂BOSS焊缝在线堆焊修复相控阵超声检验技术

核电厂反应堆换料水池与乏燃料水池冷却和处理系统（PTR）及设备循环冷却系统（RRI）中含有大量管座接头（BOSS）焊缝,其安全性和可靠性直接影响所存储核燃料的安全状态,对其进行缺陷排查和在线修复是核电厂在役检查监督的重点和难点。本文针对BOSS焊缝在线堆焊修复的特殊要求和检验难点以及射线检验的局限性,设计了一套专用的相控阵超声探头和检验工艺,试验验证结果满足堆焊修复要求,并制订了核电厂BOSS焊缝堆焊修复无损检验的方法和在役检查监督的策略。     

控制棒驱动机构石墨密封组件设计与密封性能验证

控制棒驱动机构（CRDM）检修及堆芯换料时都需要多次拆卸CRDM的耐压壳体,为解决现有耐压壳体Ω密封焊缝泄漏以及不能多次拆卸的问题,本文采用螺母压紧石墨环方案,利用石墨环的压缩回弹性能防止冷却剂泄漏,并设计一种石墨环密封组件实现快速拆卸;通过开展密封环压缩回弹测试、应力松弛测试以及密封组件泄漏率等密封性能测试试验对石墨环密封组件的密封性能进行验证。结果表明,本文设计的石墨环密封环组件满足设计要求,可以实现高温高压环境下的密封性能。      

CEPR机组CRDM役前检查方案优化

某CEPR机组的控制棒驱动机构（CRDM）耐压壳安装完成后，发现此批次CRDM焊接见证件试验存在不合格样品。为缩短CRDM更换工期，降低对于项目整体进度的影响，在对CRDM耐压壳更换过程中通过深入研究役前检查规范，提出了一种新的役前检查策略。经实践表明，采用优化后的役前检查方案，在10 d内即完成了全部CRDM的离线役前检查，较最初的计划提前了约20 d；通过对安装后的部分CRDM进行超声和涡流检查，发现离线和在线检查结果一致，并且在线检查不存在可达性问题。      

反应堆压力容器CRDM管座设计改进

对控制棒驱动机构(CRDM)管座压力试验后管座内径、垂直度等关键配合尺寸变大问题,基于目前采用的CRDM管座设计结构和制造工艺,从CRDM管座焊缝结构设计、不锈钢材料特性、管座设计强度、压力试验实施等方面对CRDM管座内径尺寸变化和垂直度变化原因进行分析,确定了尺寸变化的原因,并提出效控制关键参数的措施。     

控制棒驱动机构线圈电流曲线数据分析算法研究与实现

控制棒驱动机构(CRDM)用于实现对反应堆功率的调节与控制,其运行可靠程度直接影响着核电厂的安全性和经济性。目前核电厂和设备制造厂均缺少对CRDM进行可视化、智能化的故障诊断系统或数据分析方法。针对以上情况开发了一种控制棒驱动机构线圈电流曲线分析算法,用于对CRDM的线圈电流进行数据分析和故障诊断,从而可以实现对CRDM的智能分析和系统健康状态评估,以便于在核电厂调试及大修期间对CRDM进行全面的检修和维护。     

M310核电机组BOSS焊缝作业辐射防护控制

BOSS焊缝为管道与支管管座的连接焊缝,核电厂放射性系统管道BOSS焊缝处进行分支与变径导致容易沉积高辐射水平粒子,另由于检修空间有限,其相关作业具有很高的辐射风险。文章结合BOSS焊缝作业主要步骤,分析了作业过程存在的外照射风险与放射性污染风险,并根据各辐射风险提出切实可行的辐射防护措施。作业过程中的经验反馈和实践,为M310核电机组BOSS焊缝作业的辐射防护控制提供参考。     

Long-term installations of the DC-potential drop method in four nuclear power plants and the accuracies thereby obtained for monitoring of crack initiation and crack growth

After a total monitored operational timescale of almost five years on long-term installations, both in the laboratory and in four nuclear power plants, evidence can be put forward that the DC-potential drop method is now, at its current stage of development, suitable for inspecting and monitoring material regions such as, e.g. weld seams in pipework, for crack initiation and crack growth at power plant temperatures. This function can be performed with reliability and high sensitivity. The inspection and monitoring of cracks on the internal surface of the pipework can also be carried out from the external surface. The studies have shown that the method is basically able to monitor the growth of cracks found at discontinuous intervals using permanently installed potential probes, i.e. from plant inspection to plant inspection, while a transition to continuous monitoring is possible at any time. Thus a measure of redundancy can be provided for conventional ultrasonic and radiographic inspection, in particular for difficult to check austenitic weld seams. The method can also be seen as an alternative to the conventional techniques. When necessary, the cracks found can be measured more accurately than was previously possible with conventional ultrasonic and radiographic inspections. The total exposure to radiation can be reduced in comparison to other methods of inspection.     

Long-term installations of the DC-potential drop method in four nuclear power plants and the accuracies thereby obtained for monitoring of crack initiation and crack growth

After a total monitored operational timescale of almost five years on long-term installations, both in the laboratory and in four nuclear power plants, evidence can be put forward that the DC-potential drop method is now, at its current stage of development, suitable for inspecting and monitoring material regions such as, e.g. weld seams in pipework, for crack initiation and crack growth at power plant temperatures. This function can be performed with reliability and high sensitivity. The inspection and monitoring of cracks on the internal surface of the pipework can also be carried out from the external surface. The studies have shown that the method is basically able to monitor the growth of cracks found at discontinuous intervals using permanently installed potential probes, i.e. from plant inspection to plant inspection, while a transition to continuous monitoring is possible at any time. Thus a measure of redundancy can be provided for conventional ultrasonic and radiographic inspection, in particular for difficult to check austenitic weld seams. The method can also be seen as an alternative to the conventional techniques. When necessary, the cracks found can be measured more accurately than was previously possible with conventional ultrasonic and radiographic inspections. The total exposure to radiation can be reduced in comparison to other methods of inspection.     

Apparative developments for inservice inspections of reactor pressure vessels

Inservice inspections of primary circuit components are important preventive measures to guarantee nuclear power plant integrity, satisfying simultaneously reactor safety and economy in plant operation. Emphasizing pressurized water reactor pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implementation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system.     

Pressure vessel integrity and weld inspection procedure

The primary objective of this study is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an interrelation between nuclear pressure vessel weld integrity and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. The basic input information on rate of generation and development of weld flaws of different sizes and types is drawn primarily from published British and German studies taken almost exclusively from welds of non-nuclear pressure vessels. The input information is varied to reflect differences in weld quality and uncertainty of input data. A modified Markov process is employed and a computer code written to obtain numerical results. If it is assumed that the quality of nuclear reactor welds are the same as the quality of non-nuclear welds (i.e. the data base), then, based on the limitations of the model, the predicted critically sized defect concentration is about 50 × 10⁻⁷ per weld at the end of weld life for welds under both high and low stress if ASME, Section XI, In-Service Inspection Requirements are applied. Based on the British data and the less stringent inspection standards (compared to Section XI) the estimated number of critically sized defects per weld at the end of weld life is 250 × 10⁻⁷ and 170 × 10⁻⁷ per weld for high and low stressed welds, respectively. If it is assumed that the nuclear reactor pressure vessel welds have superior quality to the non-nuclear welds, then the model predicts an appropriately lower probability of critical defects at the end of weld life. A variety of other sensitivity studies are included in the report. Also, a simple methodology to provide an optimal weld inspection program which is consistent with a minimum cost criteria is outlined. It should be noted that the results of this study are based on the limitations of the simple model that was used and on a variety of corresponding assumptions.     

进口民用核安全设备安全检验工作中的设备审查

在进口核安全设备的安全检验中,对这些设备进行审查可以避免有缺陷的或不能证明其满足相关标准的设备用于我国核电厂,从而保障核电厂的安全运行。本文介绍了安检工作的目的、流程、内容和审查范围,重点介绍了对安检工作中设备文件的审查依据,提出了安检工作中设备文件的审查要点。     

主管道环焊缝横向拉伸试验要求和验收准则的标准对比

主管道是核电厂反应堆冷却剂系统的主动脉。某制造厂在主管道预制资质取证模拟件制作过程中出现环焊缝横向拉伸试验结果不满足RCC-M标准规范要求,通过对比不同标准规范下的管道环焊缝横向拉伸试验要求和验收准则,得出从严要求的监管原则。从而得出核电标准与规范的编制是核电国产化的关键,是核电发展实现系列化、标准化和规范化的基础,我国核电建设亟需建立一套适应国情的、统一完整的压水堆核电厂标准体系。     

核电厂核岛压力容器气压试验的方法研究

法系核电厂核岛压力容器根据在役检查规范和大纲的要求需要实施定期水压试验，但部分容器由于系统设计的原因不能用液体实施水压试验，只能执行气压试验。本文对比分析了国内外规范对于气压试验的实施要求，并结合核岛安装阶段的气压试验过程，选定了核岛压力容器气压试验的试验压力、试验介质、验收标准等；同时结合容器水压试验的风险分析和辐射防护要求，制定气压试验的防护措施。根据以上试验参数与风险防护措施，在某核电厂核岛成功实施了压力容器气压试验，为后续的在役阶段核岛压力容器气压试验提供重要参考。