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.     

Effects of thermal load and cooling condition on weld residual stress in a core shroud with numerical simulation

Stress corrosion cracking (SCC) in the heat affected zone is the primary damage form due to weld residual stress, corrosion and neutron irradiation environment in the core shroud of a boiling water reactor. The distribution of weld residual stress around a weld is necessary to be clarified to evaluate the structural integrity of core shroud for SCC. Moreover, studying the effects of welding parameters on residual stress on reducing the residual stress is very important to suppress the initiation and propagation of SCC.In this paper, we used a finite element method (FEM) to clarify the distribution of weld residual stress around the sixth horizontal weld (H6a) between the lower ring and the cylinder in the core shroud. The simulation results of axial stress were consistent with the experimental results at the inside and outside surfaces of the core shroud, respectively. The effects of thermal loads and cooling conditions were also investigated with the same model. We simulated the welding progress with water cooling on the inside and outside surfaces of the core shroud in order to study the influence of cooling conditions on the residual axial stress around the weld. The simulation results indicated that water cooling decreased the residual axial stress at the same side due to changing the temperature-affected fields. Moreover, with fixing the peak temperatures of weld passes, the simulation results of the distribution of residual axial stress by the thermal loads with different heating time were compared. The simulation results suggested that the heating time was expected to be longer and the heat flux to be smaller for reaching the small tension residual axial stress or even compression stress around the H6a weld.     

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.     

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates da/dt demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K_I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.     

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.     

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.     

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。     

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.     

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.     

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。     

Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

Various methods are being used to expand heat transfer tubes into the thick tubesheets of nuclear steam generators. The residual stresses in the as-expanded tubes and methods for reducing these stresses are important because of the role which residual stresses play in stress corrosion cracking and stress assisted corrosion of the tubing. Of the various expansion processes, the hydraulic expansion process is most amenable to analytical study. This paper presents results on the residual stresses and strains in hydraulically expanded tubes and the tubesheet as computed by two different finite element codes with three different finite element models and by a theoretical incremental analysis method. The calculations include a sensitivity analysis to assess the effects of the expansion variables and the effect of stress relief heat treatments.     

Combination of simulation and experiment in designing repair weld strategies: A feasibility study

Numerical simulations, based on an off-the-self commercial finite element (FE) code and experimental tests using the neutron diffraction (ND) technique, are combined in an attempt to evaluate post-weld heat treatment (PWHT) of a letterbox-type repair weld, in respect of its effect on the residual stress field. 21/4CrMo steel plates with an 18-pass repair weld were heat treated at various temperature levels and for different durations. Due to the prohibitive cost of a complete residual stress mapping, ND tests were performed only at selected specimen locations. In this sense, FE simulation acts as a supplement to ND, since it predicts the complete residual stress field. Uncoupled quasi-static thermoelasticity in conjunction with an element activation/deactivation technique, simulating deposition of new weld material, are combined in a 3D FE analysis. Grouping of the 18 weld beads in lumps, following a sensitivity analysis, reduces computational costs to feasible levels, whereas a creep strain hardening law is used to simulate stress relaxation during PWHT. Computed residual stresses are compared to ND measurements for verification purposes. Comparison of heat treated specimen measurements to heat treated and untreated specimen predictions illustrates that PWHT has a strong effect on the residual stress field, achieving significant stress relaxation.     

The evaluation of intergranular stress corrosion cracking problems of stainless steel piping in Taiwan BWR-6 nuclear power plant

Taiwan BWR-6 Kuosheng Nuclear Power Plant Unit 1 implemented the inspection of the intergranular stress corrosion cracking (IGSCC)-susceptible weldments of stainless steel piping in the reactor recirculation, reactor water clean-up, residual heat removal, core spray and feedwater systems. The purpose of this paper is to present the status of the fracture problems in the weldments. The crack growth analysis due to IGSCC and the standard weld overlay design based on the ASME Code Section XI and NUREG-0313 Rev. 2 for the fracture weldments are discussed in detail. Then, the contingent programs including the inspection program, fracture evaluation, and the standard weld overlay, are completely established to prevent pipe break during the reactor operation.     

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。     

The effects of environment,orientation and strength level on the stress corrosion behavior of U-0.75 wt % Ti

The effects of the environment, specimen orientation and the alloy strength level on the susceptibility of U-0.75 wt % Ti to stress corrosion crack propagation have been determined. The data show that water (H₂) is the species responsible for cracking and the Cl^? concentration has little effect on the cracking behavior. The orientation of extruded specimens had little effect on the cracking behavior. The data also show that as the strength level of the alloy increases due to varying the aging treatment the threshold stress intensity for stress corrosion crack propagation decreases linearly.     

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.     

The effect of a single tensile overload on stress corrosion cracking growth of stainless steel in a light water reactor environment

It has been found that a single tensile overload applied during constant load amplitude might cause crack growth rate retardation in various crack propagating experiments which include fatigue test and stress corrosion cracking (SCC) test. To understand the affecting mechanism of a single tensile overload on SCC growth rate of stainless steel or nickel base alloy in light water reactor environment, based on elastic-plastic finite element method (EPFEM), the residual plastic strain in both tips of stationary and growing crack of contoured double cantilever beam (CDCB) specimen was simulated and analyzed in this study. The results of this investigation demonstrate that a residual plastic strain in the region immediately ahead of the crack tips will be produced when a single tensile overload is applied, and the residual plastic strain will decrease the plastic strain rate level in the growing crack tip, which will causes crack growth rate retardation in the tip of SCC.     

Residual stresses and stress corrosion effects in cast steel nuclear waste overpacks

In the concepts for final disposal of high-level radioactive waste in Switzerland, one engineered barrier consists of an overpack made out of cast steel GS-40. Whenever tensile stresses are expected in the overpack, the issue of stress corrosion cracking must be expected. A low-strength steel was chosen to minimize potential problems associated with stress corrosion cracking. A series of measurements on stress corrosion cracking under the conditions as expected in the repository confirmed that the corrosion allowance of 50 mm used for the design of the reference overpack is sufficient over the 1000 years design lifetime. Tensile stresses are introduced by the welding process when the overpack is closed. For a multipass welding, the evolution of deformations, strains and stresses were determined in a finite-element calculation. Assuming an elastic-plastic material behavior without creep, the residual stresses are high; considering creep would reduce them. A series of creep tests revealed that the initial creep rate is important for cast steel already at 400°C.     

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.