A case study on detection and sizing of defects in steam generator tubes using eddy current testing

Eddy current testing (ECT) method is widely used to detect various types of defects occurring in nuclear steam generator tubes. Therefore, the reliability of its detection and sizing accuracy for defects should be validated. For this purpose, two tubes with defect signals were pulled from an operating steam generator and destructively examined. The defect type was a circumferential crack for one tube and an intergranular attack (IGA) for the other tube. The plus point coil probe showed a better capability to detect and size both a circumferential crack and a volumetric IGA than pancake and bobbin coil probe. The destructive results are correlated with the ECT results obtained during the in-service inspection.     

Development and Systematic Application of Surface Inspection Methods Used for In-Service Inspection of Reactor Pressure Vessel

Both from the mechanics of fracture and from actual instances of defects observed in reactor pressure vessels, it is indicated that greater importance should be attached to surface than to internal defects in the in-service inspection of these components.

In the JPDR, the reactor pressure vessel has undergone ISI three times since 1968, with emphasis placed on surface inspection, and using both remote bore scope and remote Smeck methods. The two methods gave the same results on all three occasions, so that both methods can be considered effective.

With the remote Smeck method, the lower limit of detectable defect size was found to be less than ±lmm, and the reproducibility of defect position, measured on a mock-up nozzle was better than ±3 mm. For the bore scope a newly devised “shadowing technique” is described, which appreciably enhances its ability to detect and to accurately observe surface defects.

The results of ISI in JPDR indicate that significant improvements in detection ability and accuracy could be expected by systematic application of ISI methods.     

控制棒检测中对显示信号判定技术的改进

针对控制棒涡流检测中出现的一个显示信号不能充分判定,制作了含有人造裂纹的试验样件,采用在涡流检测线圈中加入磁芯的磁饱和线圈,消除控制棒上可能存在的磁性影响。并将3点对中的涡流探头组件改进成6点对中,改善了检测条件,提高了检测的可靠性。试验研究和检测结果表明,在现有技术的前提下,穿过式线圈及点式涡流线圈均可检测出周向和轴向裂纹;穿过式线圈不能区分单个或多个裂纹,多个点式线圈存在实现这种区分的可能性;轴向裂纹的涡流信号明显,但结构信号可能会影响对周向裂纹的判伤。     

An experimental result on flaw sizing of steam generator tube at nuclear power plants using ultrasonic testing

A steam generator at a nuclear power plant consists of thousands of thin tubes, and is a highly important component in operation. Also, steam generator tubes play a critical role in maintaining pressure boundaries of the primary and secondary sides, and can be easily damaged due to operation conditions caused by high temperature and pressure. Therefore, considerable amount of efforts are being committed to evaluating structural integrity of steam generators during in-service inspection. Eddy current testing is the commonly used inspection technique to evaluate a steam generator tube's integrity, but it has limitations in accurately sizing flaws due to the nature of the technique which determines size based on the entire volume of a flaw. In this study, experiments were performed by using ultrasonic testing instead of eddy current testing for the inspection of steam generator tubes to detect various kinds of flaws and to see if the detected flaws can be sized accurately. Consequently, the ultrasonic testing technique could detect various types of flaw, and the flaw sizing results were reliable in length and depth.     

核电厂围板螺栓超声检测技术研究

围板螺栓是核电厂堆内构件的关键连接部件，长期服役下可能产生辐照应力腐蚀裂纹（IASCC）等缺陷，有必要对其结构完整性进行无损检测。分析围板螺栓的结构特点和在役检查工况，开发针对外六角头结构螺栓的组合晶片超声检测方案，介绍探头设计选型原则和缺陷评定技术，确保良好的声场有效覆盖以及检测出螺栓不同部位的裂纹缺陷。通过对含缺陷试块的试验验证了超声检查工艺的可行性，结果表明该技术能够有效检测30%螺杆横截面当量的裂纹缺陷，信噪比可达12 dB以上，满足在役检查要求。      

UT digitized data processing for in service inspection of pressurized water reactor vessels

Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection (Gagnor and Levy, 1993, Proc. 7th Asian-Pacific Conference on Nondestructive Testing, Shangaï, China, 867 pp.). The developments carried out in collaboration with the French Atomic Energy Commission (CEA) to improve the characterization of flaws detected in the body of the vessels or in the nozzles, in the vicinity of the inner or the outer surfaces now have application throughout the CIVAMIS software. The processing modules of CIVAMIS, which are implemented on site since 1994 and used by INTERCONTROLE during the in service inspections of the French PWR vessels, allow full characterization of these specific flaws. The first module is devoted to the characterization of defects located near the outer surface of the vessel or the bottom head welds (OSD module). It includes the modeling software MEPHISTOMIS which predicts the echoes coming from the interaction between the ultrasonic beam and the defects. The second module of CIVAMIS (inner surface defect module called ISD), applied to the analysis of flaws expected near the inner surface of the vessels, has been used during performance demonstration exercises on qualification mock-ups, and also on-site in five expert appraisals since its qualification in 1995. The third module available on the system has been developed and qualified for the analysis of flaws likely to appear near the inner surface of the nozzles. This module, named ‘Undercladding Crack Defect' (UCD) module, provides the operators with a set of pre-defined processing configurations well adapted to the characteristics of the transducers and of the digital acquisitions triggered in this examination case. The last new module (SAFE-END module) has been developed for the evaluation of defects located close to the bimetallic weld in the nozzles. The capacity of CIVAMIS to be adapted to each examination area and the characterization tools included in the different modules are developed in the present paper.     

The integrity of the pressure containment of the high-temperature reactor

The safety concept for ensuring the integrity of the pressure retaining containment is determined by the structural and system-specific inherent safety characteristics and features of the high-temperature reactor. The integrity of the pressure retaining containment, i.e. the elimination of a major failure, is achieved by a system of measures ensuring a high standard of quality and safety. The fundamental cornerstones of this safety concepts are the stringent requirements in the design and manufacture in view of an optimized production technology as well as specific structural solutions such as, e.g., the prestressed concrete reactor vessel. Additional safety measures such as the quality control performed independently of the manufacturer's works and the in-service inspection, have to be considered as redundant safety measures. The in-service inspection can be limited to the confirmation of safety-relevant data and analysis of deviations from these data. Recurrent non-destructive tests within the PCRV are not required, however, possible to a certain extent.     

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.     

Quantification of the reliability required of non-destructive inspection of PWR pressure vessels

This paper presents a method of quantifying the reliability required of non-destructive inspections of PWR pressure vessels. It gives a strategy for improving the effectiveness of ultrasonic non-destructive testing in assuring the integrity of a PWR vessel and allows targets of inspection reliability to be set in order to achieve the requisite level of vessel integrity. To do this the failure rate of PWR pressure vessels is predicted on the basis of a probabilistic fracture mechanics model. We use various models of the reliability of non-destructive inspection to discover the minimum level of reliability which is consistent with the desired integrity of the structure and to demonstrate how improvements can be made most effective.The reliability of inspection is usually modelled by a function giving the probability of leaving an unacceptable defect in the vessel. This function B(a) is really the “unreliability” of inspection and so 1 - B(a) gives the usual reliability. A reliable inspection is one which detects and correctly classifies defects according to some criterion usually based on size. A reliable inspection must use a technique which is intrinsically capable of detecting and sizing defects in the required size range and it must be reliably applied in practice.We find that, based on certain stated assumptions, that an inspection reliability of 80% of detecting and correctly sizing a defect of 15 mm through-wall extent yields a predicted failure rate of 10⁻⁷ per vessel year. The failure rate includes a frequency of a major accident such as a large loss of coolant (LOCA) of frequency 10⁻⁴ per vessel year. The predicted failure rate can be reduced to 10⁻⁸ per vessel year if the sizing accuracy of the technique is improved so that the chance of undersizing a 15 mm defect falls from 0.19 to about 0.01. However, the failure rate of the vessel is not predicted to decrease further with any subsequent improvement in sizing accuracy unless there is also an improvement in the asymptote of the reliability of inspection. This asymptote is due to factors beyond the capability of the technique such as, for example, human error.     

Summary report on new regulatory codes and criteria and guidelines: IAEA guidelines for WWER reactor coolant system integrity assessment

In the frame of the nuclear safety programme to assist the countries of Central and Eastern Europe, the IAEA identified and ranked in total 263 safety issues for WWER-440/230, WWER-440/213 and WWER-1000/320 nuclear power plants, related to both design and operation. In the area of reactor coolant system integrity, 24 safety issues were identified and 15 of them ranked as of high safety significance. These include: reactor pressure-vessel integrity and related aspects, primary and secondary high-energy piping integrity, steam generator integrity and reliability of the non-destructive testing for in-service inspection. In addition to obtaining international consensus on possible solutions to address the safety issues identified and to reviewing completeness of proposed safety improvements, the IAEA initiated development of guidelines to address the issues of highest safety concern. In the area of the reactor coolant system integrity, guidelines for the leak before break concept application, for the pressurized thermal shock analysis and for the in-service inspection systems qualification were developed. Further activities of the IAEA were focused on the implementation of guidelines developed in the Member States concerned. With this objective, a Co-ordinated Research Programme ‘Round-robin Exercise on WWER-440 Reactor Pressure Vessel Embrittlement, Annealing and Re-embrittlement’, a ‘WWER-440/213 Pressurized Thermal Shock Analysis Benchmark Exercise’ and a pilot study to implement the qualification approach to a real power plant component have been initiated by the IAEA and are well under way at present. In this paper, an overview of these IAEA activities related to reactor coolant system integrity is provided and the main principles and elements of guidelines developed discussed.     

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software.     

Artificial neural networks in CT–PT contact detection in a PHWR

In a pressurized heavy water reactor (PHWR), contact between calandria tubes (CT) and pressure tubes (PT) makes them susceptible to delayed hydrogen cracking. Periodic inspection of the channels must be carried out to detect the contact. As the number of channels in a PHWR is very large (306 in a 230 MW plant) periodic in-service inspection of all the channels leads to an unacceptable downtime. A non-intrusive technique that employs a system identification method is presently used for contact detection. The technique tends to overpredict the number of channels in contact, i.e. they diagnose many channels as contacting while the channels are in fact not in contact. This puts a large number of healthy channels on the at risk list reducing the efficacy of the method. This paper demonstrates the power of artificial neural networks (ANNs) in diagnosing the CT–PT contact. A counterpropagation neural network consisting of a Kohonen layer and a Grossberg layer has been employed. The noise tolerance of the network has been demonstrated.     

Inspection qualification programme for VVER reactors and review of round robin test results

Experience obtained, especially from in-service inspections of VVER 440-type reactor pressure vessels and from the Czech round test trials with international participation of ultrasonic teams, has highlighted the need for an in-service inspection qualification programme in the Czech Republic focused on NDT procedures, equipment and personnel. Recently, several national and international regional projects included in the PHARE programme (projects 4.1.2/93 and 1.02/94), briefly described, have been initiated. These projects are to cover step by step the programme of the in-service inspection qualification in view of technical justification as well as of practical assessment—performance demonstration—for all the main VVER-type primary circuit components.     

Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined.US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.     

蒸汽发生器换热管电磁超声纵向导波换能器设计与优化

高温气冷堆蒸汽发生器换热管特殊的螺旋结构导致传统外置型电磁超声导波换能器难以进行有效检测。本文针对蒸汽发生器不锈钢换热管的缺陷检测,开发了一种新型内置型电磁超声纵向导波换能器,建立了有限元多物理场耦合模型,研究了换能器铁磁结构的静态磁场分布,并对换能器激励出的纵向导波进行了时域仿真。结果表明：采用挤压聚磁的换能器结构可保证线圈附近的垂直方向磁场远大于水平方向磁场,能高效地在管道内部激励单一模式的纵向导波;优化后的探头可检测直径为5 mm的通孔缺陷和长×宽×深为20 mm×1.5 mm×1.2 mm的环向槽缺陷。因此,新型电磁超声纵向导波换能器可有效激励纵向导波,并有望应用于高温气冷堆蒸汽发生器换热管的在役缺陷检测。     

A method for quantifying the reliability of instrumentation for engineered safety features to improve accident sequence recognition in reactor control rooms

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to quantitatively evaluate the reliability of the instrumentation for engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel through the conversion of system fault trees to alarm trees. In the alarm tree, possible states of each instrumented alarm are identified as “true” or “false”. In addition, a “warning” status is also defined and integrated into the alarm analysis routine. The impact of this additional status condition on the Boolean laws used to evaluate the alarm trees is examined. An application is described for a BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents.     

Japanese regulatory practices on licensing, inspection, operational supervision and related standards

Japanese view on the safety of nuclear power plants is based on the concept that the primary responsibility for securing safety lies on electric power companies, installers of reactors.Under this concept, the Ministry of International Trade and Industry (MITI), in the course of designing and construction, has been performed an examination of the basic design and the detailed design of nuclear power plants, and in each stage of construction, a pre-operational inspection process. In addition, MITI, in operating stage, has been made throughgoing investigations on the causes of troubles and incidents as well as accidents that may affect operation, forcing utilities to take measures to prevent recurrence, and implementing safety regulation based on the “preventive maintenance” including elaborate checkings and overhaulings at the periodical inspections conducted for a period of three to four months after every 12-month operation cycle under the laws and regulations.This paper discusses the current status of nuclear power development in Japan, safety regulatory systems, views on safety and future prospects of securing safety.     

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.     

Ultrasonic signal characteristics by pulse-echo technique and mechanical strength of graphite materials with porous structure

Ultrasonic testing (UT) is an important non-destructive method to detect internal flaws and is widely applied to product control in industrial fields. In an investigation on ultrasonic signal characteristics in porous ceramics, the present authors developed an ultrasonic wave propagation model for the pulse-echo technique by improving an existing one for the transmission technique. A wave-pore reflection process was taken into account in the improvement. In the developed model, both diffusion and scattering losses can be treated as important factors of ultrasonic wave attenuation. The model was demonstrated by experimental data on ultrasonic signal characteristics of nuclear grade graphite. As an application of the model, the authors proposed a new approach combined UT signal with fracture mechanics to evaluate the mechanical strength of porous ceramics from UT signal. The combined approach was tried to apply to the acceptance test and the in-service inspection conditions of graphite components in the High Temperature Engineering Test Reactor (HTTR) as an example. This paper presents the developed propagation model for the pulse-echo technique as well as the combined approach. Moreover, both acceptance test and in-service inspection techniques of graphite components in high temperature gas-cooled reactors (HTGRs) using the combined approach was also proposed in this paper.