Risk-informed inservice inspection

The operational readiness and functional integrity of certain safety-related piping and associated structural elements such as piping supports are vital to the safety of operations in nuclear power plants. Inservice inspection (ISI) is one of the mechanisms used by the power plant owners to ensure piping integrity. Previously, the type and frequency of ISI have been based on the collective best judgment of the NRC and industry in a consensus code and rulemaking process. The ASME code-based ISI requirements and practices have not explicitly taken into consideration unique aspects of piping functions, piping degradation mechanisms, weld integrity, fabrication details, and the extent of the contribution to overall plant risk. Due to the general nature of the ASME code ISI requirements and non-reliance on quantification of risk estimates, current ISI requirements may unnecessarily emphasize inspection of less safety-significant piping segments, and thereby unnecessarily expose plant personnel to radiation exposure. Nuclear power plant owners are currently interested in optimizing inspection and testing by applying resources in more safety-significant areas. They are also interested in maintaining system availability and reducing overall maintenance costs which do not have any adverse effects on safety. The NRC has confirmed its intent of using probabilistic, as an adjunct to deterministic, techniques, to help define the scope, type, and frequency of ISI. The development of risk-informed inservice inspection programs (RI-ISI) has the potential to optimize the use of NRC and industry resources and to continue assuring adequate protection of the public's health and safety. Currently there are two methodologies being proposed by the industry for the implementation of the RI-ISI programs. One is being developed jointly by the ASME Research and Westinghouse Owners Group (WOG) and the other by the Electric Power Research Institute (EPRI). Both methodologies will be implemented for pilot plant applications. Based on discussions with the interested licensees, the NRC staff has tentatively accepted Surry, ANO-2, and Fitzpatrick as the RI-ISI pilot plants. The Surry pilot application is based on the WOG methodology, whilst ANO-2 and Fitzpatrick are based on the EPRI methodology.     

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。     

Risk-informed inservice inspection program

The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for operating nuclear power plants provides an alternative method for selecting and categorizing piping components that are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed process may support a decrease in the number of inspection and inspection intervals but will also identify areas where increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method developed by NRC staff.     

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.     

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。      

Repair and replacement of reactor internals for plant life extension

Recent experience from early Swedish BWRs corroborate that all components in a nuclear power plant can be repaired or replaced with new ones. Oskarshamn 1 has gone through a thorough refurbishment project. A number of internals were repaired or replaced including the core shroud support which was welded to the bottom of the reactor pressure vessel. The project verifies that it is fully possible to carry out complicated inspection and repair work inside a nuclear pressure vessel which has been in operation for more than 20 years. Along with increased capacity factor, operating nuclear power plants get the financial conditions needed for extensive repair and modernization projects. Large power output leads to short pay-back times for the investments. The FENIX project at Oskarshamn 1 is such a project. There are utilities whose policy is to keep their plants in as-new condition for an unlimited length of time.     

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。     

Risk assessment in setting inservice inspection priorities at light water reactors

As part of the Nondestructive Evaluation Reliability Program, sponsored by the U.S. Nuclear Regulatory Commission, Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. The method first uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The acceptable level of risk from structural failure for important systems and components is then apportioned as a small fraction of the total PRA estimated risk for core damage. This process determines the target (acceptable) risk and failure probability values for individual components. The Surry Unit 1 Nuclear Power Station was selected for pilot applications of the method. The specific systems addressed are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants.     

Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges

The paper presents a risk-informed in-service inspection (RI-ISI) pilot study project of 300 mm piping at Ignalina nuclear power plant (INPP) RBMK-1500 reactor, located in Lithuania. The RI-ISI study investigates optimal 300 mm piping ISI strategies with respect to risk and required resources. In total 1240 stainless steel welds were analyzed, assuming inter-granual stress corrosion cracking (IGSCC) to be the main damage mechanism. Pipe break frequency was estimated by probabilistic fracture mechanics methods and combined with safety barriers, provided by probabilistic safety assessment (PSA) study.After 3 years of operation, updating of RI-ISI was performed by taking into account new statistical data on pipe defects. Comparison with previous RI-ISI program was performed. The paper includes discussion on uncertainties in the study and robustness of RI-ISI programs.     

Experience and status on primary system components after 15 years of operation in the nuclear power plant obrigheim

At Obrigheim the first large pressurized light water reactor built in Germany is operating with a nominal power of 345 MW. Since the beginning of electricity production in later 1968 the nuclear power plant Obrigheim (KWO) has proved a reliable, a safe and also an economical operation with a high availability (83%) over 15 years.KWO has shown that it is possible to prove and maintain the safety and reliability of the primary components on the basis of the present regulations and safety requirements. This was achieved by careful maintenance and by applying improved non-destructive test methodsThe reactor pressure vessel with one circumferential weld near the core could be qualified for future operation by means of inservice inspection, irradiation programs, and by implementation of technical changes for normal as well as for abnormal conditions. To maintain the reliability of the steam generators, extensive eddycurrent testing of the tubes has been performed every year. In order to reduce the corrosive attack on the tubes the secondary water chemistry was controlled very sensitively by minimizing the leakage through the condensor and by using all volatile treatment. The intergranular corrosion of the tubes above the tube sheet could be reduced strongly; but an increasing number of small leakages occurred in the tube sheet region. 458 tubes had been plugged in the old steam generators before they were replaced in 1983 by new ones.In summary it can be stated that the continuous effort to maintain a high quality status of the components is responsible for the high operation availability of the plant.     

Evaluation of the proximity effect on flow-accelerated corrosion

Flow-accelerated corrosion (FAC) is a degradation mechanism that affects carbon steel piping in power plants. The failures and degradation due to FAC have necessitated numerous replacements in many power plants. Several computer codes around the world were developed as part of a systematic program or process to control FAC in power plant utilities. The typical plant model requires the input of the flow parameters, piping configuration and the plant water chemistry. The results on FAC rate are considered the key to proper selection of components for inspection. The lack of information on the effect of the upstream components located in the proximity limited the accuracy of the FAC prediction tools and hence will affect the accuracy in identifying potential inspection locations. In the present study 211 inspection data for 90° carbon steel elbows from several nuclear power plants were used to determine the effect of the proximity between two components on the FAC wear rate. The effect of the velocity as well as the distance between the elbows and the upstream components is discussed in the present analysis. Based on the analyzed trends obtained from the inspection data, significant increase in the wear rate of approximately 70% on average is identified to be due to the proximity.     

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。     

Transport of large components in Germany – some experiences and regulatory aspects

Abstract

After decommissioning of nuclear facilities, it is very often necessary to transport large components such as steam generators or reactor pressure vessels in public areas. In Germany, such shipments were carried out in 2007 and 2008 as follows: steam generators from the nuclear power plant (NPP) in Stade to Studsvik/Sweden by road and sea, reactor pressure vessel from the NPP in Rheinsberg to the interim storage facility near Greifswald by railway and steam generators from the NPP in Obrigheim to Greifswald, as well by road and inland waterway. The paper describes the experiences with these shipments, including radiation dose assessments to transport workers and the main aspects of the applied regulatory procedure by special arrangement, for which the Federal Office for Radiation Protection is the competent authority in Germany. A high level of safety was achieved for all the involved modes of transport (road, rail, sea and inland waterways). Based on these experiences, some regulatory aspects are discussed, which include classification issues of large components within the current International Atomic Energy Agency Transport Regulations, the safety concept and the use of special arrangements for such shipments and options for the further development of the International Atomic Energy Agency Transport Regulations to achieve more specific and internationally harmonised conditions or requirements for shipments of large components.     

Plant life management activities of LWR plants in Japan

Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants.     

船用核动力装置专家系统技术研究

以船用核动力装置为研究对象 ,建立了运行在仿真机上的全工况核动力装置运行仿真系统。利用智能专家控制理论 ,建立了能够管理整个装置运行的 ,能对典型运行故障进行检测与诊断的管理运行专家系统。此系统在出现故障时能及时调用知识库专家知识进行专家推理 ,分析核动力装置控制实际运行中典型故障产生的原因及解决方法、故障诊断处理具有实时性 ;利用神经网络理论 ,建立了神经网络故障检测与诊断专家系统 ,此系统将不断检测核动力装置各系统 ,并根据检测数据给出故障诊断结果。结果表明 ,在核动力装置三层智能控制结构下 ,利用神经网络故障检测与诊断专家系统对船舶核动力装置可能的运行故障进行险测与诊断 ,利用运行管理专家系统对核动力装置进行运行管理 ,提高了船用核动力装置的运行性能。所进行的专家系统研究对船用核动力装置的安全运行具有一定的指导意义     

核动力厂物项安全分级中的剂量准则研究

核动力厂的设计中通过对物项进行安全分级，来确保物项的设计、制造、建造等满足适当的要求，达到与其执行的功能相符的可靠性。本文简要介绍了根据安全重要性对核动力厂物项进行安全分级的方法以及应考虑的因素。针对物项安全分级中应考虑的未能执行某一安全功能的后果，使用核动力厂不同工况下对公众和工作人员的剂量准则来划分“高、中、低”后果。通过研究提出放射性“高、中、低”后果定量化的建议，以使得该方法在用于核动力厂物项安全分级时更具有可操作性。     

福岛核事故后核电厂安全改进行动分析

介绍了福岛核事故后世界上主要核电国家相继开展的核电厂安全检查、再评价行动,并得出相应的检查和测试结论。法国、美国和中国等国家分别提出了福岛核事故后改进核电厂安全的建议、要求和行动,并制定了具体工程措施：在极端外部事件的设防,严重事故预防和缓解,水、电、通风实体改进,限制严重事故下的放射性释放和应急准备等主要方面开展的安全改进行动,将会提高核电厂的安全水平并提升缓解严重事故的能力。反思福岛核事故,总结福岛核事故对核电安全技术改进的促进作用,对未来核电安全技术的发展进行了展望。     

Status and safety aspects of the 300 MW(e) GCFR demonstration plant

The Gulf General Atomic concept of the gas-cooled fast breeder reactor (GCFR) utilizes development that has already taken place on the high temperature gas-cooled reactor (HTGR), principally in plant systems and components, and it will benefit directly from the development work carried out to support the LMFBR — primarily the nuclear fuel development. Recent progress in the development work and engineering design for the 300 MW(e) GCFR demonstration plant is highlighted and the safety aspects of this plant are discussed.     

Decreasing snubber inservice inspection costs through snubber reduction and improved test limits

Snubber inservice inspection (ISI) requirements, along with a history of snubber malfunctions, has made inspection and maintenance of snubbers a significant part of a nuclear power plant's ISI budget. These expenses can be minimized through snubber reduction and the use of improved test limits for snubber functional testing. This paper presents a snubber overview and reviews snubber ISI requirements. Examples are given of the high cost that maintaining a snubber in an operating nuclear plant represents.Snubber reduction refers to reducing a plant's snubber population by eliminating snubbers shown not to be required to restrain piping for design basis dynamic loadings, and by replacing snubbers with other types of restraints, such as rigid struts. Snubber reduction is discussed in terms of what makes removing snubbers practical along with approaches to, and results of recently implemented snubber reduction programs.Improved or increased test limits for snubber functional testing are discussed along with an approach to, and results of an Electric Power Research Institute sponsored program to develop improved limits that would not significantly affect piping response. Improved piping acceptance criteria can be used to justify the use of increased test limits provided by snubber manufacturers. An additional use is to justify the operability of piping on which faulty snubbers were found.     

核电厂操纵员认知可靠性研究模型的选择与实验

本文采用核电厂模拟器作为研究平台 ,利用国际上流行的人的认知可靠性模型作为参考 ,运用两参数威布尔分布对三参数威布尔分布进行改造 ,建立了具有特色的中国核电厂操纵员可靠性研究理论模型 ,并应用该模型对核电厂操纵员可靠性进行了深入研究 ,与国外同类研究成果进行了比较 ,得到了一致的结果 ,该研究的进行可对核电厂的安全运行起到有益作用。