核电厂CRDM耐压壳焊缝超声检验工艺设计

控制棒驱动机构（CRDM）耐压壳属于核电厂主回路,其连接焊缝是整个放射性回路压力边界的薄弱环节,其安全性和可靠性直接影响反应堆的安全运行状态。针对CRDM耐压壳焊缝附近空间狭小、壁厚薄、可达性差等特点,本文采用仿真技术设计了一套专用的扁平型双晶聚焦超声探头和检验工艺,试验验证结果满足规程要求,解决了核电厂在役检查的监督难点,并获得了核电厂主回路Ⅰ级部件类似焊缝检验的工艺设计和验证方法。      

反应堆压力容器接管与筒体焊缝超声轮廓二维成像算法研究

核电厂复杂几何形状焊缝的超声信号位置直观显示对缺陷判断具有重要参考价值。核电厂反应堆压力容器（RPV）接管内表面通常带有一定的倾斜角度,采用传统的矩形B扫描成像算法,接管与筒体焊缝超声B/C扫描成像显示存在显示不直观、缺陷定位不准确的突出问题。本文提出了直线与直线、圆弧与椭圆弧通过圆弧相切连接的算法,利用绘制直线、弧的库函数实现轮廓的绘制并将超声信号显示在轮廓之中,形成了带轮廓的B扫描图像。通过计算在轮廓中穿过闸门线的A扫描信号的阈值,形成马鞍面形状的C扫描图像。核电厂RPV接管与筒体焊缝现场超声扫查数据验证了该算法的有效性和实用性。      

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。     

核电厂设备超声检测显示性质判定准则的对比与分析

超声检测是核电厂设备制造、建安以及在役检查活动中重要的无损检测方法,超声显示的性质判定与缺陷验收密切相关.通过对比分析我国各类核电堆型适用的设计建造和在役检查规范标准,调研了相关工业实践,选出可操作性较强、适用性较好的超声显示性质判定准则,并通过试验验证了该判定准则的准确性.     

核电站关键部件在役检查可靠性分析

基于可靠性概率统计模型和超声检测数值模型,对在役检查的可靠性进行计算与分析.以核电厂反应堆压力容器环焊缝超声检测为例,计算不同检测参数下环焊缝中裂纹类缺陷及横孔的检出率曲线和95％置信下限.结果表明,可靠性分析方法的引入可以实现在役检查工艺方法及结果的定量评估.     

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。      

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。     

秦山第二核电厂控制棒检查、分析与优化管理

由于国内核电厂控制棒运行经验少,且没有控制棒更换的相关法规或标准,为掌握控制棒包壳管状态,秦山第二核电厂通过超声、涡流等无损检测方法对在役的控制棒包壳进行了检查,得到控制棒运行可靠的技术数据,为调整和更换控制棒组件提供依据。本文通过控制棒组件典型缺陷机理分析与评价,提出了核电厂控制棒优化管理的几项措施,可以为其他核电厂控制棒管理提供重要的参考与借鉴。     

核电厂主泵安全端焊缝体积检测的改进分析

部分核电厂使用的主泵在泵壳与主管道之间存在安全端。通过采用超声和射线2种检测技术进行研究,分析比对了法国的《压水堆核岛机械设备在役检查规则》(RSE-M)与美国机械工程师协会(ASME)规范等的相关内容,并就工程实践因素进行了论述,论证了超声相比于射线在技术必要性、规范符合性、实施便利性等方面的优势,证明超声可对主泵安全端焊缝最普遍发生、最危险的裂纹进行准确、持续、经济的在役检查监督。     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

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.     

Influence of pre/in-service inspections and tests on the reliability of reactor pressure vessels

The influence of the following actions on the probability of brittle failure of the reactor pressure vessels will be estimated by probabilistic fracture mechanics: ultrasonic inspection of the welds; hydro-test of the vessel; and crack growth by normal, upset and test conditions. Taking into account that the in-service inspections and tests are done at short intervals the reliability can be shown to be extremely high.     

The evaluation of SCC defects of steel pipe using a high-energy X-ray CT sc anner

The X-ray computed CT scanner, capable of producing sharp tomograms, was expected to become a practical and revolutional means in nondestructive inspection in the industrial field.In Japan, the development of the Linac X-ray CT scanner is under way for the inspection of SCC in welds of a nuclear power plant. For development of the Linac-CT, the preliminary experiments for the inspection of SCC artificial defects were performed using a 420 kVp industrial X-ray CT scanner (TOSCANER 4200) which had been developed by Toshiba Corporation. This paper includes the background of this program and the summary of preliminary experiments for X-ray CT.     

Numerical study of defect in end plug weld joint of FBTR fuel pin

The fuel pins of fast reactor encounter high pressure, high temperature, high neutron fluence and sodium environment in their life time. This demands for high level of end plug weld integrity and hence, the acceptance criteria for these welds are very stringent. The defects usually encountered in the weld joints are micro-cracks, porosity, lack of fusion, root-pocket, etc. Amongst these weld defects, root-pocket at the weld joint is the most prominent defect observed in FBTR fuel fabrication campaign. The present paper deals with the FEM analysis of the weld zone having different sizes of root-pockets for the pressure developed inside the pin due to fission gas release and creep damage due to higher operating temperature.     

Application of eddy current inversion technique to the sizing of defects in Inconel welds with rough surfaces

This paper evaluates the applicability of eddy current inversion techniques to the sizing of defects in Inconel welds with rough surfaces. For this purpose, a plate Inconel weld specimen, which models the welding of a stub tube in a boiling water nuclear reactor is fabricated, and artificial notches machined into the specimen. Eddy current inspections using six different eddy current probes are conducted and efficiencies were evaluated for the six probes for weld inspection. It is revealed that if suitable probes are applied, an Inconel weld does not cause large noise levels during eddy current inspections even though the surface of the weld is rough. Finally, reconstruction of the notches is performed using eddy current signals measured using the uniform eddy current probe that showed the best results among the six probes in this study. A simplified configuration is proposed in order to consider the complicated configuration of the welded specimen in numerical simulations. While reconstructed profiles of the notches are slightly larger than the true profiles, quite good agreements are obtained in spite of the simple approximation of the configuration, which reveals that eddy current testing would be an efficient non-destructive testing method for the sizing of defects in Inconel welds.     

The effect of flaw orientation on the integrity of PWR pressure vessel at the events of pressurized thermal shock

The reactor pressure vessel (RPV) is the most critical component in nuclear power plants, housing the reactor core and serving as a part of the primary system pressure boundary. Because of its proximity to the reactor core, the RPV is subjected to high fast neutron flux, losing ductility and fracture toughness. At the events of pressurized thermal shock (PTS), highly embrittled RPV may not have a sufficient safety margin for fast fracture. The US NRC PTS rule requires that the reference temperature (RT_PTS) should be limited to ensure sufficient safety margins against PTS. RT_PTS=270°F was defined as the screening criterion for axial welds based on extensive quantitative evaluation of associated risks. For circumferential welds, a technical margin of 30°F was added to account for the effects of flaw orientation without same level of quantitative analysis. In this paper, the validity of the technical margin for circumferential welds is examined by comparing the quantitative risks depending on the flaw orientation. First, the result of the original work on axial welds was reproduced. Then, the risk associated with circumferential flaws was evaluated at the identical condition except for flaw orientation. The difference in screening criteria due to flaw orientation was at least 55°F, suggesting that current PTS screening criteria for circumferential flaws do not represent the same level of associated risks as that for axial flaws.