Caution when applying eddy current inversion to stress corrosion cracking

This study evaluates the applicability of computer-aided eddy current inversion techniques to the profile evaluation of stress corrosion cracking in Inconel welds. Welded plate specimens, which model head penetration welds of pressurized water reactors, are fabricated; notches and stress corrosion cracks are artificially introduced into the weld metal of the specimens. Eddy current inspections are performed using a uniform eddy current probe driven at frequencies of 10 and 40 kHz. Since weld noise is observed uniformly along the weld line, a simple signal processing is applied to eliminate it. First, the artificial notches are reconstructed and good agreements between reconstructed and true profiles are provided, which demonstrates that the computer-aided eddy current inversion technique can deal with defects in welds. Then, numerical simulations are performed to evaluate the profiles of the stress corrosion cracks. In the numerical simulations, the stress corrosion cracks are modeled as a conductive region with a fixed width of 0.3 mm. The cross-sectional profiles of the cracks are reconstructed from measured eddy current signals directly above and along a crack. Although eddy current signals calculated from the reconstructed profiles agree well with measured ones, the true profiles revealed by destructive testing are found to be very different from the reconstructed ones. Whereas the most plausible reason for the difference is the unexpectedly volumetric profile of the stress corrosion cracks, this study has revealed that computer-aided eddy current inversion techniques that have been used to consider cracks in thin structures would not at this point be directly applicable to those in thick structures. It is also important to know in advance those crack features that can adversely impact accurate crack sizing including whether a detected crack is volumetric or not, namely there are many parallel cracks in a cluster or not.     

Evaluation of stress corrosion cracking as a function of its resistance to eddy currents

This study discusses the equivalent conductivity, the equivalent width, and the equivalent resistance of stress corrosion cracks from the viewpoint of eddy current testing. Four artificial stress corrosion cracks were prepared for this study, and their eddy current signals were gathered using two absolute pancake probes and two differential type plus point probes. Then their numerical models were evaluated using finite element simulations on the basis of the measured eddy current signals and their profiles revealed by destructive tests. The results of this study revealed that whereas the equivalent conductivity and the equivalent width depend on the exciting frequency utilized, the equivalent resistance of a crack has much less dependency, which agrees well with an earlier report. This study also revealed that the resistance of a crack depends on probe utilized. Larger probes tend to lead to smaller crack resistance. Pancake type probes tend to lead to larger crack resistance than plus point probes. Analyzing the results together with earlier reports indicates that cracks with a large equivalent conductivity tend to have large equivalent width, and supports the validity of assuming the minimum resistance of a stress corrosion crack whereas considering the conductivity and the width individually would not be viable.     

Loading conditions in horizontal feedwater pipes of LWRs influenced by thermal shock and thermal stratification effects

In recent years, a particular form of crack formation has occurred in a number of pressurized and boiling water reactors on the internal surfaces of horizontal lengths of feedwater piping upstream of the steam generators and reactor pressure vessels.The fractographic evaluation and the orientation of the cracks show that these are to be attributed to cyclic stressing in the axial direction. Comparison of the stresses due to thermal shock and thermal stratification reveals that, on account of the associated load cycles, the cracks were in essence caused by thermal stratification. This is also indicated by the orientation of the cracks.The present results of the corrosion tests show that with high oxygen content (450 ppb) and temperature level (210°C) the strain rate at thermal stratification exerts an essential influence on the number of cycles to crack initiation. With the conservative test conditions described, the values may fall below those given in the ASME fatigue design curve. In the case of strain rates that apply to thermal shock, the cycles to crack initiation are on the safe side of the curve.The remedial measures taken by KWU by the installation of the siphon reduce the frequency of stress amplitudes. It can be concluded from corrosion tests simulating these conditions that with the strain rates occurring in this case, the number of cycles to crack initiation are on the safe side according to the ASME fatigue design curve.     

Surveillance of defects under cyclic loads and during hot and cold hydro-tests by acoustic emission, UT and potential drop techniques (German HDR-programme)

In the framework of the German HDR reactor safety programme a selected area of the reactor pressure vessel was loaded with 9500 thermal cycles. In 3 saw cut regions several crack fields were generated and monitored with on-line and off-line NDE-procedures. With potential drop and ultrasonic testing, crack front profiles could be described off-line with sufficient resolution. During the trials the crack growth was measured with stationary potential drop and ultrasound probes. With on-line acoustic emission crack growth of about 5 mm could be detected. On the basis of an interpretation model a classification of the acoustic emission signals in crack growth and crack surface friction phenomena was possible. The crack growth during the cooling phase is caused by fatigue; the reason for crack growth during the heating up phase is corrosion influence. All cracks are growing at different time periods and have different stress conditions.     

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.     

Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment

Advanced transmission electron microscopy techniques were carried out in order to investigate stress corrosion cracking in Alloy 600 U-bend samples exposed in simulated PWR primary water at 330 °C. Using high-resolution imaging and fine-probe chemical analysis methods, ultrafine size oxides present inside cracks and intergranular attacks were nanoscale characterized. Results revealed predominance of Cr₂O₃ oxide and Ni-rich metal zones at the majority of encountered crack tip areas and at leading edge of intergranular attacks. However, NiO-structure oxide was predominant far from crack tip zones and within cracks propagating along twin boundaries and inside grains. These observations permit to suggest a mechanism for intergranular stress corrosion cracking of Alloy 600 in PWR primary water. Indeed, the results suggest that stress corrosion cracking is depending on chromium oxide growth in the grain boundary. Oxide growth seems to be dependent on oxygen diffusion in porous oxide and chromium diffusion in strained alloy and in grain boundary beyond crack tip. Strain could promote transport kinetic and oxide formation by increasing defaults rate like dislocations.     

核级304L不锈钢钎焊接头组织及耐腐蚀性能研究

核级304L不锈钢与BNi 7钎料真空钎焊接头存在晶间腐蚀行为,但工艺与钎缝耐腐蚀性能的关系尚未得到充分研究。为充分评估压水堆燃料组件结构件中不锈钢真空钎焊接头的晶间腐蚀和应力腐蚀敏感性,降低腐蚀失效风险,采用定量金相方法分析了钎缝中的化合物相含量,采用硫酸 硫酸铁法和双环动电位再活化（DL EPR）法评价了钎缝耐晶间腐蚀性能,并采用高温高压水应力腐蚀裂纹扩展试验评价了钎缝的耐应力腐蚀性能。结果表明,钎缝中化合物相含量越高,耐晶间腐蚀性能越好。且钎缝在高温高压水中存在明显的应力腐蚀开裂行为,但其与钎焊工艺的关系尚需进一步试验研究。     

Effect of shot peening on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an operating PWR plant

The effect of shot peening on the primary stress corrosion cracking behavior of thermally treated Alloy 600 steam generator tubes in an operating pressurized water reactor (PWR) plant was analyzed based on pulled tube examinations and in-service inspection eddy current test (ISI-ECT) data. The evaluation was focused on the shape of crack, evolution of the number of new cracks and cracked tube fraction, and variation of crack length and the corresponding eddy current amplitude before and after shot peening. The shape of the crack was changed from a half-elliptical type before shot peening, to an elliptical one with bulging after peening. It was concluded that the shot peening was not effective for retarding both crack initiation and growth for this plant.     

Towards an elastic-plastic fracture mechanics predictive capability for reactor piping

Intergranular stress corrosion cracks have been discovered in the recirculation bypass piping and core spray lines of several boiling water reactor (BWR) plants. These cracks initiate in heat-affected zones of girth welds and grow circumferentially by combined stress corrosion and fatigue. Reactor piping is mainly type 304 stainless steel, a material which exhibits high ductility and toughness. A test program described in this paper demonstrates that catastrophic crack growth in these materials is preceded by considerable amounts of stable crack growth accompanied by large plastic deformation. Thus, conventional linear elastic fracture mechanics, which only applies to the initiation of crack growth in materials behaving in a predominantly linear elastic fashion, is inadequate for a failure analysis of reactor piping.This paper is based upon research initiated by a need to develop a realistic failure prediction and a way to delineate leak-before-break conditions for reactor piping. An effective engineering solution for the type of cracks that have been discovered in BWR plants was first developed. This was based upon a simple net section flow stress criterion. Subsequent work to develop an elastic-plastic fracture mechanics methodology has also been pursued. A survey of progress being made is described in this paper. This work is based on the use of finite element models together with experimental results to identify criteria appropriate for the onset of crack extension and for stable crack growth. A number of criteria have been evaluated. However, the optimum fracture criterion has not yet been determined, even for conditions which do not include all of the complications involved in reactor piping.     

Modeling of SCC initiation and propagation mechanisms in BWR environments

During operation of mainly BWRs’ (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 °C – corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 °C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with increasing global stresses. In accordance with respective experimental literature data it is shown that decreasing chloride and increasing pH levels of the primary bulk water at 288 °C reduce the total crack propagation rates including anodic path corrosion as well as hydrogen cracking. It is also demonstrated that crack propagation rates can be significantly suppressed by hydrogen water chemistry (HWC) that leads to reduction of bulk surface corrosion potentials. As a conclusion the extended SSC-model for nickel supplies quantitative insight into the frequently controversially discussed high temperature SCC mechanisms of a basic alloying element of BWR components.     

The experimental investigation of a crack's influence on the concrete breakout strength of a cast-in-place anchor

Actual model tests are carried out to investigate the influence of a crack on the concrete breakout strength of an anchorage and to examine the appropriateness of the various criteria that are available for an anchorage design in a new nuclear power plant and the anchorage capacity evaluation in the currently operating nuclear power plants in the case where various types of cracks already exist in the anchorage. These tests are intended for the cast-in-place anchor which is widely used for the fastening of equipment in Korean nuclear power plants. One hundred and five plain concrete test specimens are manufactured on the basis of ASTM E488 containing 21 cracked conditions. The principal test variables chosen are related to the crack properties including, the crack width, the crack depth and the distance between the crack and the anchor. In order to apply a tensile force, a 100 tonnes-capacity actuator is used with a displacement control of 0.5 mm/min. It is proven that the ACI 349 code, CEB-FIP code and SQUG GIP are conservative in a comparison with the test results. Some improved proposals are suggested on the basis of the test results in order to accurately take into account the influence of the crack in the anchorage capacity evaluation.     

The effect of high temperature reactor primary circuit helium on the formation and propagation of surface cracks in alloy 800 H and inconel 617

The depths of surface cracks caused by the combined effects of corrosion and creep strain have been measured in specimens of Inconel 617 (Ni-22Cr-9Mo-12Co-1A1) and Alloy 800 H (Fe-32Ni-20Cr). Tests were carried out at 1073–1223 K in impure helium simulating the primary circuit coolant of a high temperature gas-cooled reactor, and in air. A characteristic crack depth $\text{a}{}_{90}$ was derived to represent the observed crack length distribution. In both test environments, an incubation strain was observed for the formation of surface cracks. Although the presence of surface cracks led to an increase in the creep rate, the magnitude of the change was similar for the two test atmospheres. Finally, a method has been developed to allow the depth of surface cracks, characterized by $\text{a}{}_{90}$, to be plotted with the stress rupture curves in order to indicate whether the corrosion effects need to be considered in the derivation of design stresses.     

AL-6XN合金在超临界水中的应力腐蚀研究

采用慢拉伸试验研究了AL-6XN合金在550~650℃、25 MPa超临界水中的应力腐蚀行为,使用扫描电镜观察了材料断口形貌与标距面裂纹分布。结果显示:550℃试验条件下材料表现为穿晶开裂,标距面裂纹很多且分布较均匀,试样边角处出现扇形河流花样;650℃试验条件下材料表现为沿晶开裂,断口呈现典型的冰糖状,标距面裂纹数量大幅减少且集中于断口附近,相应的延伸率和断面收缩率也大幅降低。质子辐照对AL-6XN宏观力学性能影响不大,但导致试样断口的扇形花样(起裂源)数量增加,解理台阶宽度增大。以上结果表明,AL-6XN在超临界工况下具有严重的应力腐蚀开裂敏感性,升高温度和质子辐照会明显提高材料的应力腐蚀敏感性。     

Effect of cladding on the stress intensity factors in the reactor pressure vessel

In general, reactor pressure vessels (RPV) are cladded with stainless steel to prevent corrosion and radiation embrittlement. The ASME Sec. XI specifies that a subclad crack which may be found during the in-service inspection must be considered as a semi-elliptical surface crack when the thickness of cladding is less than 40% of the crack depth. In order to refine the fracture assessment procedures for such subclad cracks, three-dimensional finite element analyses were applied for various subclad cracks embedded in the base metal. A total of 18 crack geometries were analyzed, and the results were compared with those for idealized semi-elliptical surface cracks for two different loading conditions, i.e. internal pressure and pressurized thermal shock. The resulting stress intensity factors for subclad cracks were 50–70% less than those for idealized surface cracks. It has been proven that the condition specified on the ASME Sec. XI is overly conservative for subclad cracks which are assumed to be surface cracks.     

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.     

Primary water stress corrosion cracking behaviors in the shot-peened Alloy 600 TT steam generator tubings

The effect of compressive residual stress on the primary water stress corrosion cracking behavior was investigated, based on the J-1 and J-2 nuclear power plant data. The following analyses were performed such as: (i) Weibull slope; (ii) crack growth rate; (iii) average crack length; (iv) crack length distribution. Alloy 600 TT exhibits strong heat to heat variations in its sensitivity to PWSCC. Crack growth rate was retarded after shot-peening. The compressive residual stress induced by shot-peening was more effective on new, short cracks, than on existing, long cracks. However, whether the ‘new’ cracks were initiated after peening is an unresolved issue, due to the present ECT sensitivity limit.     

Crack propagation in a LMFBR elbow

This paper provides results from a large scale elbow fracture mechanics fatigue test at LMFBR operating temperature and with sodium as fluid. The material, a Type 304 stainless steel, is one of the selected stainless steels used in breeder reactor design. Crack initiation, crack shape development, ligament instability and safety margins against gross plastic instability were predicted. Initial cracks were located at the crown of the elbow where the highest stresses occur under in-plane bending, which was the loading condition for the test. The nearly pure bending stress across the wall is regarded as a typical loading in LMFBR structures. Cracks under such unfavourable stress distribution extend preferably lengthwise before wall penetration, as compared with cracks under membrane stresses.The experiment, conducted at stress levels approaching the maximum design values, demonstrates low crack growth rates under plant conditions, showing that crack extension during service would be quite small. In fact, more than 28 times the expected transients would be required to advance a crack of 3 mm depth and 30 mm length to penetrate the wall in the region of highest stress.     

Determination of the critical crack length for steam generator tubing based on fracture-mechanics-based method

The steam generator (SG) tubing, as a key ingredient in the primary coolant circuit, is the weakest link that affects the availability and safety of a nuclear power plant. For safety reasons, it is very important to predict the life span of SG tubing. The critical crack lengths at different stages should be determined when the life span can be predicted. In this article, the critical crack lengths at different stages are determined in the form of graphs based on fracture-mechanics approach. From these graphs, it is concluded that the critical crack lengths for fatigue, stress corrosion cracking (SCC), and rupture are 0.11 mm, 0.36 mm and 7.20 mm respectively under accident conditions, and 1 mm, 2 mm, 43 mm respectively under normal conditions. It can also be concluded that the crack propagation mechanism can be divided successively into three or four stages, namely the corrosion stage, the fatigue stage (if the load is turbulent), the SCC stage and the rupture stage. Finally, some advices for the accelerated life test are given.     

Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Würgassen Nuclear Power Station need not be feared. Although the material and corrosion–chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert's agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept.     

新锆合金碘致应力腐蚀性能研究

研究了再结晶状态的相同成分，不同织构取向的两块新锆合金板材在不同的实验条件下的抗碘致应力腐蚀性能。用直流电位降法动态监测裂纹的长度并对试样进行了织构的测定、第二相的观察和断口分析结果表明，在实验条件下，N18-2(L-T)抗碘致应力腐蚀的能力优于N18-1(T—L)断口形貌表明，在应力腐蚀裂纹的初始扩展阶段，断口沿晶开裂；在裂纹的稳态扩展阶段，以穿晶准解理扩展为主。