Residual stress prediction of dissimilar metals welding at NPPs using support vector regression

Residual stresses are an important factor in the component integrity and life assessment of welded structures. In this paper, a support vector regression (SVR) method is presented to predict the residual stress for dissimilar metal welding according to various welding conditions. Dissimilar welding joint between nozzle and pipe is regarded in the analyses since it has been known to be highly susceptible to Primary Water Stress Corrosion Cracking (PWSCC) in the primary system of a nuclear power plant (NPP). The residual stress distributions are predicted along two straight paths of a weld zone: a pipe flow path on the inner weld surface and a path connecting two centers of the inner and outer surfaces of a weld zone of a pipe. Four SVR models are developed for four numerical data groups which are split according to the two end section constraints and the two prediction paths and the SVR models are optimized by a genetic algorithm. The SVR models are trained by using a data set prepared for training, optimized by using an optimization data set, and verified by using a test data set independent of the training data and the optimization data. It is known that the SVR models are sufficiently accurate to be used in the integrity evaluation by predicting the residual stress of dissimilar metal welding zones.     

核电厂蒸汽发生器传热管与管堵头的一次侧应力腐蚀及其防护

一次侧应力腐蚀（PWSCC）是一种晶间腐蚀，是因敏感的管子微观结构、高的残余拉应力和工作应力以及腐蚀性环境（高温水）引起的。防止PWSCC的措施包括：选择适当的管子材料、减小残余拉应力和改善腐蚀性环境、激光焊接衬管以及镀镍修补。     

Analysis of welding distortion due to narrow-gap welding of upper port plug

Narrow-gap welding is a low distortion welding process. This process allows very thick plates to be joined using fewer weld passes as compared to conventional V-groove or double V-groove welding. In case of narrow-gap arc welding as the heat input and weld volume is low, it reduces thermal stress leading to reduction of both residual stress and distortion. In this present study the effect of narrow-gap welding was studied on fabrication of a scaled down port plug in the form of a trapezoidal box made of 10 mm thick mild steel (MS) plates using gas tungsten arc welding (GTAW). Inherent strain method was used for numerical prediction of resulting distortions. The numerical results compared well with that of the experimentally measured distortion. The validated numerical scheme was used for prediction of weld induced distortion due to narrow-gap welding of full scale upper port plug made of 60 mm thick SS316LN material as is proposed for use in ITER project. It was observed that it is feasible to fabricate the said port plug keeping the distortions minimum within about 7 mm using GTAW for root pass welding followed by SMAW for filler runs.     

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.     

Influence of Dissolved Hydrogen on Structure of Oxide Film on Alloy 600 Formed in Primary Water of Pressurized Water Reactors

In order to investigate the relationship between the susceptibility of primary water stress corrosion cracking (PWSCC) in Alloy 600 and the content of dissolved hydrogen (DH) in the primary water of pressurized water reactors (PWR), structural analysis of oxide films formed under four different DH conditions in simulated primary water of PWR was carried out using a grazing incidence X-ray diffractometer (GIXRD), a scanning electron microscope (SEM) and a transmission electron microscope (TEM). In particular, to perform accurate analysis of the thin oxide films, the synchrotron radiation of SPring-8 was used for GIXRD.

It has been observed that the oxide film is mainly composed of nickel oxide, under the condition without hydrogen. On the other hand, needle-like oxides are formed at 1.0 ppm of DH. In the environment of 2.75 ppm of DH, the oxide film has thin spinel structures. From these results and phase diagram consideration, the condition around 1.0 ppm of DH corresponds to the boundary between stable NiO and spinel oxides, and also to the peak range of PWSCC susceptibility. This suggests that the boundary between NiO and spinel oxides may affect the SCC susceptibility.     

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.     

Reduction method for residual stress of welded joint using harmonic vibrational load

Welding is widely used for construction of many structures. Since welding is a process using locally given heat, residual stress is generated near the bead. Tensile residual stress degrades fatigue strength. Some reduction methods of residual stress have been presented and, for example, heat treatment and shot peening are practically used. However, those methods need special tools and are time consuming. In this paper, a new method for reduction of residual stress using harmonic vibrational load during welding is proposed. The proposed method is examined experimentally for some conditions. Two thin plates are supported on the supporting device and butt-welded using an automatic CO₂ gas shielded arc welding machine. Residual stress in the direction of the bead is measured by using a paralleled beam X-ray diffractometer with scintillation counter after removing quenched scale chemically. First, the welding of rolled steel for general structure for some excitation frequencies is examined. Specimens are welded along the groove on both sides. For all frequencies, tensile residual stress near the bead is significantly reduced. Second, welding of the specimen made of high tensile strength steel is examined. In this case, tensile residual stress near the bead is also reduced. Finally, the proposed method is examined by an analytical method. An analytical model which consists of mass and preloaded springs with elasto-plastic characteristic is used. Reduction of residual stress is demonstrated using this model.     

The effect of stress levels on the coefficient of thermal expansion of a fine-grained isotropic nuclear graphite

Due to the fluctuation and non-uniform distribution of temperature within the core structure of high-temperature gas-cooled reactors (HTGRs), the thermal expansion behavior of graphite materials plays an important role in the design of graphite components, especially of large-scale components. In the present paper, in order to investigate the influence of stress levels on the coefficient of thermal expansion (CTE) of IG-110 graphite, the strain gauge method was used to measure the CTE on the cylindrical specimens under a series of loads applied using a universal tensile testing machine. In addition, a more precise measurement using a thermal dilatometer was employed to validate the tests using the strain gauge method. A good agreement has been obtained between the experimental results using these two methods. The results show that when the specimens were under compressive loads, the CTE along the loading direction of the specimens increased and that along the perpendicular direction decreased, with more changes in the former. The absolute changes of the CTE in the two directions increased with increasing applied load. When graphite specimens were subjected to a compressive load of 40 MPa, the axial CTE of specimens sectioned along the radial direction of the graphite brick as it is installed in the core structure increased from 4.13 × 10⁻⁶ to 5.35 × 10⁻⁶ K⁻¹, while the axial CTE of specimens sectioned along the vertical direction increased from 3.97 × 10⁻⁶ to 5.58 × 10⁻⁶ K⁻¹. Moreover, the residual change of the CTE, which was caused by the permanent residual strain after unloading, was observed. The change of the CTE with stress levels should be considered in the stress analysis and life prediction of the nuclear graphite components.     

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.     

2219铝合金搅拌摩擦焊和钨极保护焊焊接件残余应力的中子衍射研究

利用中子衍射法对2219铝合金搅拌摩擦焊（FSW）和钨极保护焊（TIG）焊接件开展了三维残余应力测量,并对残余应力分布规律进行了分析。结果表明：焊接件的纵向残余应力数值较大;FSW焊接件残余应力整体较TIG焊接件的小;FSW和TIG焊接件的残余拉应力最大值分别为101 MPa和174 MPa,FSW焊接件残余拉应力最大值较TIG焊接件的小;FSW残余拉应力最大值处于轴肩边缘,且前进侧峰值大于后退侧峰值;TIG焊接件残余拉应力最大值处于焊缝边缘。通过中子衍射实验获得的焊接件残余应力分布,将可用于焊接工艺的优化与焊接件的寿命预测。     

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.     

Calculation of the power peaking factor in a nuclear reactor using support vector regression models

Local power density (LPD) at the hottest part of a hot nuclear fuel rod should be estimated accurately to confirm that the rod does not melt. The power peaking factor (PPF) is defined as the highest LPD divided by the average power density in the reactor core. In this paper, the PPF is calculated by support vector regression (SVR) models using numerous measured signals from the reactor cooling system. SVR models are regression analysis models using a kernel function for artificial neural networks. Their neural network weights are found by solving a quadratic programming problem under linear constraints. SVR models are trained using a training data set and then verified against another test data set. The proposed SVR models were applied to the first fuel cycle of the Yonggwang nuclear power plant unit 3. The root mean square errors of the SVR model, with and without in-core neutron flux sensor signal inputs, were 0.1113% and 0.0968%, respectively. This level of errors is sufficiently low for use in LPD monitoring.     

Thermo-mechanical characteristics of concrete at elevated temperatures up to 310 °C

Both, the normal strength concretes (NSC) and high strength concretes (HSC) have been used in structures which may be exposed to elevated temperatures. Concretes have also been used in the construction of radiation shielding structures. Data on the behaviour of concrete at high temperature is of prime concern in predicting the constructions and safety of buildings in response to certain accidents or particular service conditions. Prediction of mechanical behaviour, thermo-mechanical deformations and moisture migration in non-uniformly heated concrete is important for safe operation of concrete containment.This paper presents the results of an experimental investigation carried out to predict the behaviour of concrete intended for nuclear applications. For this purpose, normal concrete having compressive strength of 40 MPa was designed using limestone aggregates. Cylindrical specimens (110 mm × 22 mm) were made and subjected to heating-cooling cycles at 110, 210 and 310 °C. Measurements were taken for thermal gradient, mass loss, deformations, residual mechanical properties, thermal conductivity, and porosity. This investigation developed some important data on the properties of concrete exposed to elevated temperatures up to 310 °C. Comparisons and interesting conclusions were drawn about the thermal stability at high temperature and the residual mechanical properties of the tested concrete.     

Creep rupture strength of activated-TIG welded 316L(N) stainless steel

316L(N) stainless steel plates were joined using activated-tungsten inert gas (A-TIG) welding and conventional TIG welding process. Creep rupture behavior of 316L(N) base metal, and weld joints made by A-TIG and conventional TIG welding process were investigated at 923 K over a stress range of 160-280 MPa. Creep test results showed that the enhancement in creep rupture strength of weld joint fabricated by A-TIG welding process over conventional TIG welding process. Both the weld joints fractured in the weld metal. Microstructural observation showed lower δ-ferrite content, alignment of columnar grain with δ-ferrite along applied stress direction and less strength disparity between columnar and equiaxed grains of weld metal in A-TIG joint than in MP-TIG joint. These had been attributed to initiate less creep cavitation in weld metal of A-TIG joint leading to improvement in creep rupture strength.     

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

Numerical simulation of creep damage for low alloy steel welded joint considering as-welding residual stress

A model to calculate the welding temperature and residual stress was built using finite element code ABAQUS, and a subroutine of creep damage was also developed. Based on the coupling of welding residual stress and creep damage, the welding residual stress and creep damage of a tube made of Cr5Mo steel were simulated. This method can obtain the distributions of complex residual stress, creep damage and stress relaxation, which provide a reference for discussing the effect of residual stress on creep damage. The results show that the welding residual stress is very large at initial stage, then it is relaxed in a short time at high temperature. The distribution of creep and damage is mainly decided by the as-welding residual stress. Welding residual stress has a great effect on the creep and damage, which provides a reference for the design and life prediction of high temperature component.     

Control of welding residual stress for ensuring integrity against fatigue and stress–corrosion cracking

The availability of several techniques for residual stress control is discussed in this paper. The effectiveness of these techniques in protecting from fatigue and stress–corrosion cracking is verified by numerical analysis and actual experiment. In-process control during welding for residual stress reduction is easier to apply than using post-weld treatment. As an example, control of the welding pass sequence for multi-pass welding is applied to cruciform joints and butt-joints with an X-shaped groove. However, residual stress improvement is confirmed for post-weld processes. Water jet peening is useful for obtaining a compressive residual stress on the surface, and the tolerance against both fatigue and stress–corrosion cracking is verified. Because cladding with a corrosion-resistant material is also effective for preventing stress–corrosion cracking from a metallurgical perspective, the residual stress at the interface of the base metal is carefully considered. The residual stress of the base metal near the clad edge is confirmed to be within the tolerance of crack generation. Controlling methods both during and after welding processes are found to be effective for ensuring the integrity of welded components.     

Numerical simulation of the combined effects of plasma heating and neutron heating loads on the ITER first wall

To understand the combined effect of plasma heating and neutron heating loadings, the distributions of temperature, stress, and strain in different two-dimensional first wall panel models under normal ITER operation condition were simulated using finite element method. The maximum temperature occurs at the Be armor, and reaches 461 °C. High thermal stresses (in the range of 80-200 MPa) are found at the interface between the Be armor and the CuCrZr layer. The maximum thermal stress reaches 324 MPa in the SS316L cooling tube (20 mm diameter), exceeding its yield strength and resulting in a maximum strain of about 1.7% at the tube inner surface. These simulation results are useful for the design and operation of ITER.     

Remove Welding Residual Stress for CFETR Vacuum Vessel by Trailing Ultrasonic Impact Treatment

The China fusion engineering test reactor (CFETR) vacuum vessel is welded by narrow gap TIG (NG-TIG) welding, and the welding residual stress of the CFETR vacuum vessel can be redistributed by trailing welding ultrasonic impact treatment. In order to investigate the feasibility of the residual stress removing scheme, and to obtain the optimal trailing ultrasonic impact treatment technological parameters in the process of removing welding residual stress, a welding model that similar to vacuum vessel welding seam is established by using ABAQUS software, a NG-TIG welding heat source subroutine which is written in FORTRAN language used to simulate NG-TIG welding. According to the welding simulation results, a trailing welding ultrasonic impact treatment model is established, and the effects of the impact pin number, the impact method, the impact pin diameter and the impact frequency on welding residual stress are studied. The results show that the longitudinal residual stress in welding seam and its adjacent area and the lateral residual stress in the whole region have been obviously decreased by different trailing welding ultrasonic impact processes, and have made the tensile stress in the welding seam and its adjacent area has been changed into compressive stress, which can provide theoretical guidance and reference for actual production.     

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.