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.     

Monitoring of inside surface crack growth by strain measurement of the outside surface: A feasibility study

Cracks detected by in-service inspections are not always removed when they are judged to be not hazardous according to fitness-for-service evaluations. In order to secure the integrity of the cracked components, it is important to confirm that the cracks do not grow notably beyond the growth prediction conducted for the judgement. However, due to the limitation of accuracy of size determination by the current inspection techniques such as ultrasonic testing, it is difficult to know how much the cracks have grown since their previous measurement. In this study, feasibility of a crack growth monitoring method (outside strain monitoring method) was evaluated by finite element analyses and experiments. When a pipe deforms elastically due to internal pressure, the strain at its outside surface increases. The magnitude of strain near the crack differs from that at an uncracked portion, and the difference depends on the crack size. Elastic finite element analyses were performed for cracked pipes under internal pressure for various crack sizes. It was shown that, by measuring the change in strain at the outside surface of the cracked pipe, the crack size and how much the crack grew can be identified. In the experiment, cracked pipes were subjected to static internal pressure and strains for eight cracks of different sizes were measured. It was revealed that the maximum error was 0.44 mm for the estimation of crack depth of 4 mm and 0.28 mm for the estimation of 1 mm crack growth in the depth direction.     

Control of the safe operating period of equipment and pipelines in reactor facilities which are in operation or under construction

The basic assumptions for ensuring safe operation of the components of nuclear facilities, based on controlling service lifetime characteristics, are presented. It is shown up for the Du300 RBMK-1000 pipes, which were made of corrosion-resistant austenitic steel, that this technology can be used effectively in operating power generating units. The complex of measures, developed and validated in the last few years, for monitoring and assessing the technical state of weld seams in pipelines, using the safety concepts “leak before rupture” and “prevention of failure” as well as methods for suppressing the proneness of weld seams to form cracks by the corrosion cracking method under stress, has made possible not only safe operation of Du300 pipelines but it is also a basis for optimizing the volumes and periodicity of operational nondestructive monitoring. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 61–65, July, 2007.     

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。     

In-service monitoring and fractographic inspection of ultrasonic indications on a feed water suction pipe of the Gundremmingen nuclear power plant

Radiographic and ultrasonic examinations of three pipes of the Gundremmingen nuclear power plant revealed cracks with a maximum depth of about 2 mm on the inside of the pipes with a wall thickness of 7.1 mm. For continued service until replacement two representative areas were monitored for crack growth by means of the potential drop method. As crack growth during the last year of service can be excluded by metallographic examinations after replacement the result of monitoring by means of potential probes – in relation to its resolution capacity no crack growth in the period under monitoring – is confirmed. A final evaluation based on the actual crack depths results in minimum resolution of 0.2 and 0.5 mm in crack depth. The suitability of the potential drop method for in-service monitoring of existing indications for crack growth – even until replacement of the component, if necessary – could thus be demonstrated convincingly.     

Crack detection and sizing technique by ultrasonic and electromagnetic methods

Improvements in defect detection and sizing capabilities for non-destructive inspection techniques have been required in order to ensure the reliable operation and life extension of nuclear power plants. For the volumetric inspection, the phased array UT technique has superior capabilities for beam steering and focusing to objective regions, and real-time B-scan imaging without mechanical scanning. In contrast to the conventional UT method, high-speed inspection is realized by the unique feature of the phased array technique. A 256-channel array system has developed for the inspection of weldment of BWR internal components such as core shrouds. The TOFD crack sizing technique also can be applied using this system. For the surface inspection, potential drop techniques and eddy current techniques have been improved, which combined the theoretical analysis. These techniques have the crack sizing capability for surface breaking cracks to which UT method is difficult to apply. This paper provides the recent progress of these phased array and electromagnetic inspection techniques.     

Development of simplified evaluation method for creep-fatigue crack propagation

In the design assessment of fast reactor plant components, prevention of crack initiation from defect-free structures is a main concern. However, existence of initial defects such as weld defects cannot be entirely excluded and this potential cracks are to be evaluated to determine if initiated cracks do not lead to component failure instantly. Therefore, evaluation of structural integrity in the presence of crack-like defects is also important to complement the formal design assessment. The authors have been developing a guideline for assessing long-term structural integrity of fast reactor components using detailed inelastic analysis and nonlinear fracture mechanics. This guideline consists of two parts, evaluation of defect-free structures and flaw evaluation. In the latter, creep-fatigue is considered to be one of the most essential driving force for crack propagation at high operating temperature exceeding 500 °C. The uses of J-integral-type parameters (fatigue J-integral range and creep J-integral) are recommended to describe creep-fatigue crack propagation behavior in the guideline. This paper gives an outline of the simplified evaluation method for creep-fatigue crack propagation.     

Advanced flaw production method for in-service inspection qualification mock-ups

One of the key issues in in-service inspection qualification is the representativeness of the defects used in qualification specimens. The best representativeness is achieved with realistic defects. However, present specimen production techniques have some significant weaknesses, such as unrealistic defects or additional alterations induced in the surrounding material. Specimens manufactured, for example, by weld implantation or with weld solidification defects always result in one or more extra weld interfaces. These interfaces can be detected by NDT. To overcome problems with the current specimens, a new defect manufacturing technique was developed. The new technique produces natural, representative defects without introducing additional weld metal or other unwanted alterations to the specimen.The new method enables artificial production of single, separate fatigue cracks by thermal loading. The method is based on a natural thermal fatigue damage mechanism and enables production of real cracks directly into the samples. Cracks are produced without welding or machining and without any preliminary surface treatment or artificial initiator such as a notch or a precrack. Single crack or a network of cracks can be induced into the base material, welded areas, HAZ, weld claddings, threaded areas, T-joints, etc. The location, orientation and size of produced cracks can be accurately controlled. Produced cracks can be used to simulate different types of service-induced cracks such as thermal fatigue, mechanical fatigue and stress corrosion cracks. It is shown that artificially produced thermal fatigue cracks correspond well with the real, service-induced cracks and overcome the problems of traditional qualification specimen manufacturing techniques.     

Enhancements of eddy current testing techniques for quantitative nondestructive testing of key structural components of nuclear power plants

In this paper, studies on upgrade of eddy current testing (ECT) techniques for inspection of stress corrosion cracks (SCC) in key structural components of a nuclear power plant are reported. Access and scanning vehicle (robot), advanced probes for steam generator (SG) tube inspection, developments and evaluations of new ECT probes for welding joint, and ECT-based crack sizing technique are described, respectively. Based on these techniques, it is demonstrated that ECT can play as a supplement of ultrasonic testing (UT) for the quantitative inspection of welding zone. It is also proved in this work that new ECT sensors are efficient even for inspection of a stainless steel plate as thick as 15 mm.     

Surface-crack aspect ratio development during corrosion-fatigue crack growth in low-alloy steels

The development of aspect ratios (crack depth/half-crack length), was studied for semi-elliptical surface cracks in low-alloy steel undergoing corrosion-fatigue in an elevated temperature aqueous environment. Water-enhanced crack growth behavior is influenced by the concentration of hydrogen sulfide at the crack tip, and the sulfide concentration is in turn influenced by mass-transport considerations. The mass-transport characteristics of surface cracks may be different from those of more common test specimens; e.g. compact tension specimens. It is also shown that the method of preparing surface-cracked specimens can have an influence upon the crack growth behavior; surface cracks emanating from crack-starter notches may behave differently than ‘natural’ surface cracks because of differences in the mass-transport paths. It is also shown in that the rate of water flow along the length of a surface crack can affect the resulting crack aspect ratio and crack growth rates.     

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。      

Numerical investigation of ductile crack growth behavior in a dissimilar metal welded joint

In this paper, the finite element method (FEM) based on GTN model is used to investigate the ductile crack growth behavior in single edge-notched bend (SENB) specimens of a dissimilar metal welded joint (DMWJ) composed of four materials in the primary systems of nuclear power plants. The J-Δa resistance curves, crack growth paths and local stress-strain distributions in front of crack tips are calculated for eight initial cracks with different locations in the DMWJ and four cracks in the four homogenous materials. The results show that the initial cracks with different locations in the DMWJ have different crack growth resistances and growth paths. When the initial crack lies in the centers of the weld Alloy182 and buttering Alloy82, the crack-tip plastic and damage zones are symmetrical, and the crack grow path is nearly straight along the initial crack plane. But for the interface cracks between materials and near interface cracks, the crack-tip plastic and damage zones are asymmetric, and the crack growth path has significant deviation phenomenon. The crack growth tends to deviate into the material whose yield stress is lower between the two materials on both sides of the interface. The different initial crack locations and mismatches in yield stress and work hardening between different materials in the DMWJ affect the local stress triaxiality and plastic strain distributions in front of crack tips, and lead to different ductile crack growth resistances and growth paths. For the accurate integrity assessment for the DMWJ, the fracture toughness data and resistance curves for the initial cracks with different locations in the DMWJ need to be obtained.     

恰希玛核电工程安全壳筒体钢衬里的埋弧自动焊焊接质量控制

介绍了巴基斯坦恰希玛核电工程(C-2)安全壳钢衬里筒体的基本状况,分析了埋弧自动焊产生终端裂纹的原因.防止埋弧自动焊产生终端裂纹,匹配主要工艺参数是消除焊缝宽度窄而余高大的基本控制方法,正确选择焊剂的颗粒度并合理回收使用焊剂是减小气孔及焊缝表面压气/凹坑缺陷的重要途径,通过二次切割可减小翘曲等波浪变形现象.     

Vessel behavior following a through-wall crack

A fracture mechanics model has been developed to predict the behavior of a reactor pressure vessel following the occurrence of a through-wall crack during a pressurized thermal shock event. This study has been coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory. The fracture mechanics model uses as inputs the critical transients and probabilities of through-wall cracks from the IPTS Program. The model has been applied to predict the modes of failure for plant specific vessel characteristics. A Monte Carlo type of computer code has been written to predict the probabilities of alternate failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. The computer code also calculates the crack driving force as a function of the crack length and the internal pressure for critical times during the transient. The fracture mechanics model has been applied in calculations that simulate the Oconee-1 reactor pressure vessel. The model predicted that about 50% of the through-wall axial cracks will turn and follow a circumferential weld giving a potential for missiles. Missile arrest calculations predict that vertical as well as all potential horizontal missiles will be arrested and will be confined to the vessel enclosure cavity. In future work, plant specific analyses will be continued with calculations that simulate Calvert Cliff-1 and H.B. Robinson-2 reactor vessels.     

Wave propagation visualization in an experimental model for a control rod drive mechanism assembly

Nondestructive inspection techniques such as ultrasonic testing, eddy current testing, and visual testing are being developed to detect primary water stress corrosion cracks in control rod drive mechanism (CRDM) assemblies of nuclear power plants. A unit CRDM assembly consists of a reactor upper head including cladding, a penetration nozzle, and J-groove dissimilar metal welds with buttering. In this study, we fabricated a full-scale CRDM assembly mock-up. An ultrasonic propagation imaging (UPI) method using a scanning laser ultrasonic generator is proposed to visualize and simulate ultrasonic wave propagation around the thick and complex CRDM assembly. First, the proposed laser UPI system was validated for a simple aluminium plate by comparing the ultrasonic wave propagation movie (UWPM) obtained using the system with numerical simulation results reported in the literature. Lamb wave mode identification and damage detectability, depending on the ultrasonic frequency, were also included in the UWPM analysis. A CRDM assembly mock-up was fabricated in full-size and its vertical cross section was scanned using the laser UPI system to investigate the propagation characteristics of the longitudinal and Rayleigh waves in the complex structure. The ultrasonic source location and frequency were easily simulated by changing the sensor location and the band pass filtering zone, respectively. The ultrasonic propagation patterns before and after cracks in the weld and nozzle of the CRDM assembly were also analyzed. Since this visualization method is not limited in the flat cross section, it will be useful in developing ultrasound-based structural health monitoring technologies, advanced nondestructive methods, and numerical models. In addition, the proposed laser UPI system could be a useful tool in optimizing the receiver and transmitter locations, the ultrasonic path, and the ultrasonic frequency.     

Stress corrosion crack length and depth evolution on pulled SG tubes from a PWR

Eddy current test (ECT) data collected from the in-service inspection (ISI) of pulled steam generator (SG) tubes were evaluated in terms of the primary water stress corrosion crack (PWSCC) length and depth evolution. After shot peening, the evaluated crack length and the number of cracks did not increase, but the Eddy current voltages that were related to the crack depth increased continuously. In the analysis of all the tubes of the plant, the evaluated crack lengths saturated at around 6 mm, while the voltage of the defects increased with time. As a result, shot peening was considered effective for suppressing crack length increase, but not so successful from the point of preventing crack deepening. It was also found that tube bundles that were susceptible to PWSCC were located in a special area depending on the steam generators of the analyzed plant.     

Incorporation of passive components aging into PRAs

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants.     

Formation and growth of cracking in feed water pipes and RPV nozzles

The observation of numerous small and large cracks in ferritic feed water pipes of boiling (BWR) and pressurized water reactors (PWR) in the last few years has led to basic research into the causes of cracking and the crack growth mechanisms.In horizontal feed water pipe sections connected to nozzles of reactor pressure vessels (RPV) of BWR's as well as of steam generators (SG) of PWR's, circumferential macro and micro cracks were detected. These cracking phenomena could be observed in base material of pipes as well as in weld seam regions. The examination of the stress state displayed that the cracked pipe regions have been exposed to a number of cyclic thermal transients (thermal shock, flow stratification) during start-up (hot stand-by) and shut-down periods of the plants. During thermal transient periods, local and global cyclic stresses in the referred pipe cross sections have been induced which in interaction with the influence from environment (in operation as well as in shut-down periods) and local geometrical imperfections led to the initiation and formation of macro and micro cracks.In the reactor water clean-up system of BWR through which reactor water is fed from the RPV to the main feed water line, two longitudinally welded elbows have been detected to be severely cracked. Both elbows have been subjected to an internal pressure corresponding to RPV and additionally to a relevant “in-plane” bending moment. These longitudinal cracks were found to be started from the inner elbow surface. In one case the longitudinal crack was situated in the base material and was enlarged to leakage. In the second elbow the longitudinal crack was located in the heat affected zone (HAZ) of a longitudinal weld. In both cases the macro cracks started either from corrosion pits located in defective areas of the magnetic protection layer or from geometrical notches (weld root). The semi-elliptic small cracks got linked to more extended shallow cracks.Formation and growth mechanism of these cracks have been studied at the MPA Stuttgart in laboratory under simulated operation conditions which were held as realistic as possible compared with those in nuclear power plants.The results of experimental studies in laboratory as well as conclusions based on the above mentioned cracking phenomena in piping have been used as basic information for a realistic design of large scale (RPV) thermal shock experiments under operation conditions. The formation and growth mechanism of these cracks and their detection by means of NDE during thermal transients at the inner surface of RPV nozzle and at the adjacent cylindrical areas of RPV shell will be described.     

Prediction of crack coalescence of steam generator tubes in nuclear power plants

Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory’s Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments.