Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water reactor (BWR). Therefore, the evaluation of SCC propagation is important for the safe maintenance of the core shroud. This paper attempts to predict the remaining life of the core shroud due to SCC failures in BWR conditions via SCC propagation time calculations. First, a two-dimensional finite element method model containing H6a girth weld in the core shroud was constructed, and the weld processing was simulated to determine the weld's residual stress distribution. Second, using a basic weld residual stress field, the SCC propagation was simulated using a node release option and the stress redistribution was calculated. Combined with the J-integral method, the stress intensity factors were calculated at depths of 2, 3, 4, 8, 12, 16, 19, 22, 25 and 30 mm in the crack setting inside the core shroud; then, the SCC propagation rates were determined using the relation between the SCC propagation rate and the stress intensity factor. The calculations show that the core shroud could safely remain in service after 9.29 years even when a 1-mm-deep SCC has been detected.     

Prediction of residual stress distributions due to surface machining and welding and crack growth simulation under residual stress distribution

In nuclear power plants, stress corrosion cracking (SCC) has been observed near the weld zone of the core shroud and primary loop recirculation (PLR) pipes made of low-carbon austenitic stainless steel Type 316L. The joining process of pipes usually includes surface machining and welding. Both processes induce residual stresses, and residual stresses are thus important factors in the occurrence and propagation of SCC. In this study, the finite element method (FEM) was used to estimate residual stress distributions generated by butt welding and surface machining. The thermoelastic-plastic analysis was performed for the welding simulation, and the thermo-mechanical coupled analysis based on the Johnson-Cook material model was performed for the surface machining simulation. In addition, a crack growth analysis based on the stress intensity factor (SIF) calculation was performed using the calculated residual stress distributions that are generated by welding and surface machining. The surface machining analysis showed that tensile residual stress due to surface machining only exists approximately 0.2 mm from the machined surface, and the surface residual stress increases with cutting speed. The crack growth analysis showed that the crack depth is affected by both surface machining and welding, and the crack length is more affected by surface machining than by welding.     

Stress corrosion cracking countermeasure observed on Ni-based alloy welds of BWR core support structure

The effect of hydrostatic test on the residual stress re-distribution was simulated by experiment to confirm the residual stress behavior of the cone-shaped shroud support to reactor pressure vessel (RPV) weld, where a number of cracks due to stress corrosion cracking (SCC) were observed on the inner side only. Test specimen with tensile residual stress was loaded and unloaded with axial plus bending load, which simulates the hydrostatic test load, and the strain change was measured during the test to observe the residual stress behavior. The results verify that the residual stresses of the shroud support to the RPV weld were reduced and the stresses on inner and outer sides were reversed by the hydrostatic test. As the SCC countermeasure, the shot peening (SP) technology was applied. Residual stress reduction by SP on the complicated configuration, and improvement of SCC resistance and endurance of the compressive residual stress were experimentally confirmed. Then, SP treatment procedures on the actual structure were confirmed and a field application technique was established.     

贯穿件J形坡口焊接残余应力分析

核电站反应堆压力容器（RPV）顶盖控制棒驱动机构（CRDM）管座J形坡口焊缝在一回路高温高压水环境下存在应力腐蚀开裂（SCC）的风险，而焊接残余应力是SCC的主要驱动力。使用二维轴对称模型有限元方法对CRDM中心管座J形坡口进行焊接残余应力分析。为了探索一种简单、高效和保守的方法，研究了热源简化、焊缝形状简化、屈服强度、相变和强化行为对焊接残余应力的影响。结果表明:双椭球热源与均匀热源得到的残余应力结果基本一致；焊缝形状由鱼鳞状简化为方块模型对焊接残余应力结果影响不大，但是与合并焊道的结果相差较大；采用低屈服强度得到的残余应力结果并不保守；在ANSYS软件中，固液相变对残余应力结果影响不大；等向强化模型的结果比随动强化模型的结果保守；在工程上，建议采用均匀热源、方块焊道模型和等向强化模型进行焊接模拟。      

Modeling of weld residual stresses in core shroud structures

This paper presents a computational model to predict residual stresses in a girth weld (H4) of a BWR core shroud. The H4 weld is a multi-pass submerged-arc weld that joins two type 304 austenitic stainless steel cylinders. An axisymmetric solid element model was used to characterize the detailed evolution of residual stresses in the H4 weld. In the analysis, a series of advanced weld modeling techniques were used to address some specific welding-related issues, such as material melting/re-melting and history annihilation. In addition, a 3-D shell element analysis was performed to quantify specimen removal effects on residual stress measurements based on a sub-structural specimen from a core shroud. The predicted residual stresses in the H4 weld were used as the crack driving force for the subsequent analysis of stress corrosion cracking in the H4 weld. The crack growth behavior was investigated using an advanced finite element alternating method (FEAM). Stress intensity factors were calculated for both axisymmetric circumferential (360°) and circumferential surface cracks. The analysis results obtained from these studies shed light on the residual stress characteristics in core shroud weldments and the effects of residual stresses on stress corrosion cracking behavior.     

Predicting welding residual stresses in a dissimilar metal girth welded pipe using 3D finite element model with a simplified heat source

Dissimilar metal welds are commonly used in nuclear power plants to connect low alloy steel components and austenitic stainless steel piping systems. The integrity assessment and life estimation for such welded structures require consideration of residual stresses induced by manufacturing processes. Because the fabrication process of dissimilar metal weld joints is considerably complex, it is very difficult to accurately predict residual stresses. In this study, both numerical simulation technology and experimental method were used to investigate welding residual stress distribution in a dissimilar metal pipe joint with a medium diameter, which were performed by a multi-pass welding process. Firstly, an experimental mock-up was fabricated to measure the residual stress distributions on the inside and the outside surfaces. Then, a time-effective 3-D finite element model was developed to simulate welding residual stresses through using a simplified moving heat source. The simplified heat source method could complete the thermo-mechanical analysis in an acceptable time, and the simulation results generally matched the measured data near the weld zone. Through comparing the simulation results and the experimental measurements, we can infer that besides the multi-pass welding process other key manufacturing processes such as cladding, buttering and heat treatment should also be taken into account to accurately predict residual stresses in the whole range of the dissimilar metal pipe.     

Numerical weld modeling — a method for calculating weld-induced residual stresses

In the past, weld-induced residual stresses caused damage to numerous (power) plant parts, components and systems (Erve, M., Wesseling, U., Kilian, R., Hardt, R., Brümmer, G., Maier, V., Ilg, U., 1994. Cracking in Stabilized Austenitic Stainless Steel Piping of German Boiling Water Reactors — Characteristic Features and Root Causes. 20. MPA-Seminar 1994, vol. 2, paper 29, pp.29.1–29.21). In the case of BWR nuclear power plants, this damage can be caused by the mechanism of intergranular stress corrosion cracking in austenitic piping or the core shroud in the reactor pressure vessel and is triggered chiefly by weld-induced residual stresses. One solution of this problem that has been used in the past involves experimental measurements of residual stresses in conjunction with weld optimization testing. However, the experimental analysis of all relevant parameters is an extremely tedious process. Numerical simulation using the finite element method (FEM) not only supplements this method but, in view of modern computer capacities, is also an equally valid alternative in its own right. This paper will demonstrate that the technique developed for numerical simulation of the welding process has not only been properly verified and validated on austenitic pipe welds, but that it also permits making selective statements on improvements to the welding process. For instance, numerical simulation can provide information on the starting point of welding for every weld bead, the effect of interpass cooling as far as a possible sensitization of the heat affected zone (HAZ) is concerned, the effect of gap width on the resultant weld residual stresses, or the effect of the ‘last pass heat sink welding’ (welding of the final passes while simultaneously cooling the inner surface with water) producing compressive stresses in the root area of a circumferential weld in an austenitic pipe. The computer program (finite element residual stress analysis) was based on a commercially available code (Hibbitt, Karlsson, Sorensen, Inc, 1997. user's manual, version 5.6), and can be used as a 2-D or 3-D FEM analysis; depending on task definition it can provide a starting point for a fracture mechanics safety analysis with acceptable computing times.     

焊接残余应力对316LN不锈钢应力腐蚀裂纹扩展速率的影响

核电站不锈钢管道焊接过程中引入的残余应力对焊接接头的应力腐蚀开裂性能有较大影响。本文针对一AP1000主管道316LN不锈钢焊接模拟件进行残余应力分析和应力腐蚀裂纹扩展速率测量,得到了焊后原始状态和去应力热处理状态的焊接热影响区材料在高温高压水中的应力腐蚀裂纹扩展速率。实验结果表明,焊接残余应力明显提高了热影响区的应力腐蚀裂纹扩展速率,且在含氢的压水堆一回路正常水化学下焊接残余应力的影响更加显著。     

Evaluation of stainless steel pipe cracking: Causes and fixes

This paper discusses (1) studies of impurity effects on susceptibility to intergranular stress corrosion cracking (IGSCC), (2) intergranular crack growth rate measurements, (3) finite-element studies of the residual stresses produced by induction heating stress improvement (IHSI) and the addition of weld overlays to flawed piping, (4) leak-before-break analyses of piping with 360° part-through cracks, and (5) parametric studies on the effect of through-wall residual stresses on intergranular crack growth behavior in large diameter piping weldments. The studies on the effect of impurities on IGSCC of Type 304 stainless steel show a strong synergistic interaction between dissolved oxygen and impurity concentration of the water. Low carbon stainless steel (Type 316NG) appear resistant to IGSCC even in impurity environments. However, they can become susceptible to transgranular SCC with low levels of sulfate or chloride present in the environment. The finite-element calculations show that IHSI and the weld overlay produce compressive residual stresses on the inner surface, and that the stresses at the crack tip remain compressive under design loads at least for shallow cracks.     

Intergranular stress corrosion cracking behavior of niobium-added Type 308 stainless steel weld overlay metal in a simulated BWR environment

Type 308 stainless steel weld metal as an internal cladding of reactor pressure vessels for boiling water reactors is subject to postweld heat treatment during fabrication and can suffer sensitization depending on carbon and ferrite contents. This sensitization can be avoided by using niobium-added Type 308 weld metal (specified as Type 308 NbL) which was developed for one-layer overlay welding application. In the present study, stress corrosion cracking (SCC) behavior of heat-treated Types 308 and 308NbL weld metals in oxygenated high temperature pure water was evaluated by slow strain rate test and U-bend tests with and without crevice. Every test showed that Type 308NbL weld metals were highly resistant to SCC compared to ordinary Type 308 weld metals. In single U-bend test, one-layer overlay weld metals of Type 308NbL produced by electroslag welding process using wide strip electrodes were crack free over 23,000 h. The U-bend test data of ordinary Type 308 weld metals were successfully analyzed by an SCC reaction model. Using this analysis, the SCC life margin for Type 308NbL over ordinary Type 308 weld metals, expressed as a ratio of respective times to SCC initiation, was estimated to be about 36.     

Current research on environmentally assisted cracking in light water reactor environments

The effect of dissolved oxygen level on fatigue life of austenitic stainless steels is discussed and the results of a detailed study of the effect of the environment on the growth of cracks during fatigue initiation are presented. Initial test results are given for specimens irradiated in the Halden reactor. Impurities introduced by shielded metal arc welding that may affect susceptibility to stress corrosion cracking are described. Results of calculations of residual stresses in core shroud weldments are summarized. Crack growth rates of high-nickel alloys under cyclic loading with R ratios from 0.2 to 0.95 in high-purity water that contains <5 and 300 ppb dissolved oxygen at 240, 289, and 320°C, are summarized.     

Simplified analysis of shrinkage in pipe to pipe butt welds

Generally some shrinkage is typical of butt welding of pipes. Shrinkage due to butt welding could be more pronounced and significant in thin wall stainless steel pipes because the thermal expansion coefficient is roughly one and half times that of carbon steel. An axisymmetric finite element evaluation of hoop shrinkage associated with circumferential butt welds in thin wall stainless steel pipes was performed. Actual shrinkage data for a larger (24 in. diameter, 0.375 in. wall thickness) pipe and a smaller (4 in. diameter, 0.237 in. wall thickness) pipe were utilized. The results indicate that very localized residual stresses in excess of yield strength produced during cooldown of metal in the weld and heat affected zones cause redistribution of the stresses. A simplified elastic–plastic analysis approach was developed with adjustments for section modulus and Poisson’s ratio, and the strains due to radial shrinkage were calculated for inside and outside surfaces of the pipe at the weld center line. From the strain point of view, the strain values in the circumferential direction were about 1.4% for the larger size pipe and 3.4% for the smaller size pipe. The strain values in the axial direction were 2.5% for the larger pipe and 5.9% for the smaller pipe. It is concluded that these levels of strains are not detrimental in nature. However, for the smaller pipe they are on the high side and it is recommended not to use the pipe for elevated temperature service. Residual stresses were also calculated for inside and outside surfaces of the pipe at weld center line using a simplified elastic–plastic approach and a bilinear stress–strain curve and compared with published data indicating a general agreement.     

Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals

The stress corrosion cracking (SCC) rate of reactor internals of boiling water reactors (BWR) continues to increase with on-line operating years. The recent occurrences of cracking in the weld heat affected zones of high carbon stainless steel core shrouds correlate with the years of operation and the water chemistry history. Recently, cracking has also been found in shrouds that were constructed of low carbon or stabilized stainless steels. While these steels are more resistant to intergranular stress corrosion cracking (IGSCC) in the as-fabricated condition, this field experience establishes that the conditions under which the materials will crack in core structures are attributable to the combined effects of high residual stresses, associated with the shroud construction, the presence of a more aggressive, oxidizing environment in the core and to microstructural changes in the material. These changes result from the manufacturing process as well as thermal exposure during operation. Studies of materials that have cracked in the field, as well as high temperature material studies in the laboratory, are being performed to understand the mechanisms. The use of highly oxidizing, high purity water environments is integral to reproducing the conditions for cracking. The status of the laboratory efforts to gain understanding and to verify the mechanisms are presented. Modeling of IGSCC is also a key tool used to understand the cracking behavior of the low carbon stainless steels. The PLEDGE (Plant Life Extension Diagnosis by GE) model is used to support the hypotheses that tie together the role of the different contributing elements: residual stress, core water chemistry and microstructural features. The crack growth modeling is also used to evaluate the benefits of different strategies to manage and mitigate cracking of reactor internals such as hydrogen water chemistry.     

321-52M-690异种金属焊接接头的应力腐蚀开裂行为研究

采用慢应变速率拉伸(SSRT)和高温电化学相结合的方法,研究了外加电位对321-52M-690异种金属焊接接头在含Cl-高温高压水中应力腐蚀开裂(SCC)倾向的影响规律。结果表明,在300℃、50ppm Cl-环境下,焊接接头的SCC敏感性随电极电位(-700~+100mV)的升高而增大,且存在一个介于0~+50mV(vs.SHE)之间的临界电位Ecrit。当电极电位低于Ecrit时,焊接接头的SCC敏感性较小,SCC敏感性指数ISSRT基本在40%左右,断裂形式为外力主导的塑性开裂;当电极电位高于Ecrit时,ISSRT急剧增加至70%以上,断裂形式为腐蚀主导的脆性开裂。试样断裂位置均位于硬度最低的321母材处,表明在321/690异种金属焊接接头中321母材对SCC最为敏感,故进一步探讨了321不锈钢的应力腐蚀开裂行为和机理。     

反应堆压力容器内部大尺寸环形异种金属焊缝残余应力分布研究

获得反应堆压力容器内部大尺寸环形异种金属焊缝残余应力分布可为反应堆压力容器结构设计和制造工艺优化提供指导,通过设计和制造能够代表产品焊接结构形式的镍基合金和低合金钢异种金属焊接结构模拟件,采用轮廓法测试焊接结构模拟件内部纵向残余应力,采用有限元法模拟计算焊接结构模拟件横向和纵向残余应力,获得了整个异种金属焊接接头残余应力分布特征。结果表明:焊缝区域内部纵向残余应力为拉伸应力,峰值应力达到500 MPa左右,并且表层应力大于内部应力,峰值应力出现在距下表面3 mm和24 mm位置;横向残余应力在焊缝区域从上表面到下表面的分布为拉应力-压应力-拉应力,压缩横向残余应力峰值达到?300 MPa,出现在距下表面约18 mm位置。本文研究可为焊接结构设计提供理论指导。      

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.     

控制棒驱动机构上部Ω焊缝堆焊修复结构完整性分析

针对核电厂控制棒驱动机构（CRDM）上部Ω焊缝堆焊修复（WOR）技术,采用数值模拟方法进行了修复结构完整性评估。根据堆焊修复参数制定二维轴对称高斯热源等效输入,并采用ANSYS程序的单元生死技术模拟焊接过程,得到了结构的焊接残余应力。考虑电厂运行的全部瞬态,计算了结构的瞬态应力,并开展了疲劳分析。结合焊接残余应力分析和瞬态应力分析的结果,开展了断裂力学分析。结果表明,WOR结构的疲劳结果、应力强度因子及裂纹扩展等方面均能满足相应的规范要求。      

焊后热处理对16MND5钢焊接组织结构与残余应力的影响

16MND5钢广泛应用于核岛承压容器构件,其焊接接头不可避免地会引入高的残余应力,而焊后热处理可有效消减焊接残余应力以克服应力腐蚀裂纹的影响。本工作利用轮廓法和中子衍射技术研究了焊后热处理对16MND5钢焊接残余应力的影响。结果表明,轮廓法与中子衍射测试结果在趋势和数值上取得了较好的一致性,焊后热处理使焊接态的残余应力峰值从约420 MPa降低至约210 MPa。同时,利用金相法和SEM研究了焊后热处理对焊缝区域组织结构的影响。结果表明,焊后热处理主要表现为贝氏体和少量自回火马氏体的焊缝中心组织转变为回火贝氏体和回火马氏体,热处理后的焊缝区晶粒明显长大。     

Long-Term Integrity of Waste Package Final Closure for HLW Geological Disposal, (III)

Overpack, a high-level radioactive waste package for Japan's geological disposal program, is required to isolate the sealed vitrified waste from contact with groundwater for 1,000 years. In this study, EBW (Electron Beam Welding), a typical application of high energy density beams widely used in various industries, was examined for its applicability to sealing a carbon steel overpack lid with a thickness of 190 mm. Welding conditions and welding parameters were examined for single-pass welding in a narrow area for three different penetration depths. Weld joint tests including macro- and microstructure, and mechanical properties were conducted and assessed within tentatively applied criteria for weld joints. Measurement and numerical calculation for residual stress were also conducted and the tendency of residual stress distribution was discussed. These test results were compared with the basic requirements of the welding method for overpack, which were pointed out in our first report. The induced void and cold shut inside the weld joint and surface roughness were also discussed for their improvement and evaluation, which need to be established to assure the long-term integrity of overpack lid closure.     

BWR pipe crack and weld clad overlay studies

Several topics pertaining to the problem of stress corrosion cracking (SCC) of piping in boiling water reactors are addressed in this paper: (1) the effects of impurities, dissolved oxygen content, and strain rate on susceptibility of SCC of “Nuclear Grade” Type 316NG and sensitized Type 304 stainless steel, (2) finite-element analyses and experimental measurement of residual stresses in weldments with weld overlays, and (3) analysis of field components to assess effectiveness of in-service inspection techniques and the in-reactor performance of weld overlays. Several anion impurities including sulfates, chlorides, nitrates, borates, and carbonates were studied under both near neutral and slightly acidic conditions. At the low impurity concentrations expected in reactor coolant systems (<0.1 ppm), the sulfur species appear to be the most deleterious. They promote intergranular SCC in sensitized stainless steel and transgranular SCC in the low-carbon “Nuclear Grade” stainless steel. Correlations between experimental data and a phenomenological model that describes the effect of strain rate on SCC are presented. Measurements of the residual stresses produced by weld overlays confirm that the process is very effective in producing compressive stresses on the inner surface of the weldment. Examination of a weld overlay removed from service suggests that no additional throughwall crack growth had occurred after application of the overlay.