Weld Overlay Cladding Research Articles

Optimisation of Weld Overlay Cladding Parameters Using Full-Factorial Design of Experiment

Weld Overlay Cladding (WOC) shares the same scientific principals as conventional welding where there are multiple governing factors that control the process and outcome. The present work employs a Design of Experiment (DoE) approach to optimising process parameters for cladding a nickel superalloy onto low alloy steel with the aim to improve productivity and quality. The arc current, the clad metal heating current were identified as the key process variables for this stage of experimentation. A full-factorial 4-by-2 test was carried out to identify the optimal levels. Results showed that there is a mild positive trend between the height of individual strings of beads and both variables. However no relationship was established with the depth of penetration, nor with the height of single or double layer stacks. The optimal level of the variables was therefore chosen to be the one that has the highest productivity rate as there were no significant differences. Further experimentation has been planned and described in this paper.

Influences of Cr content and PWHT on microstructure and oxidation behavior of stainless steel weld overlay cladding materials in high temperature water

Influences of Cr content and post weld heat treatment (PWHT) on microstructure and oxidation behavior of stainless steel cladding materials in high temperature water were investigated. The amounts of metal oxidized and dissolved were estimated to compare the oxidation behaviors of cladding materials with different Cr contents and PWHT. The results indicated that higher Cr content led to formation of more ferrite content, and carbides were found along δ/γ phase interface after PWHT. Higher Cr content enhanced the pitting resistance and compactness of the oxide film to reduce metal amount oxidized and dissolved, which mitigated the weight changes and the formation of Fe-rich oxides. PWHT promoted more and deeper pitting holes along the δ/γ phase interface due to formation of carbides, which resulted in an increase in metal amount oxidized and dissolved, and were also responsible for more Fe-rich oxides and higher weight changes.

Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400°C for periods of 100–10,000h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100h, and gradually for aging durations greater than 1000h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000h, had the highest number density after aging for 2000h, and coarsened after aging for 10,000h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition.

Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1×1024n/m2 at 290°C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic–plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

Development of a flexible robotic welding system for weld overlay cladding of thermoelectrical plants’ boiler tube walls

The walls of the boilers of power plants undergo corrosion and wear during operation. The reduction in thickness is measured periodically to prevent collapse of the tubes and consequent operation stoppages for long periods. The solution that is usually applied to increase the availability of power plants is gas metal arc welding (GMAW) cladding of the walls with metal alloys that are more resistant to wear. Application occurs in two distinct locations: the boiler and the workshop. Small worn areas are coated inside the boiler itself, while in the workshop, panels are coated so that they can either be used in the boiler’s most affected regions or to produce new walls with higher wear resistance. Mechanized welding produces better results than manual welding; however, in current systems, the oscillating movement of the torch is generated by a single axis. This limitation renders the welding process unstable if the aim is to reduce bead reinforcement and the number of weld beads by increasing the amplitude of the torch oscillation. This instability produces high spattering and excess penetration due to the change in the contact tip to work distance (CTWD), since the power of the welding electric circuit is a function of this distance and since the constant changes in CTWD disturb the metal transfer dynamics. This paper shows that it is possible to overcome the technological difficulties for both application sites and to make automatic cladding operations more productive and of a better quality. For this, a dedicated Computer Numeric Control (CNC) robot with four degrees of freedom was developed, as well as the methodology for generation of its trajectory. The cladding obtained by pulsed GMAW with ER309L stainless steel is presented in order to validate the developments.

Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2×1019ncm−2 (E>1MeV) and a flux of 1.1×1013ncm−2s−1 at 290°C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300°C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2×1023nm−2 (E>1MeV) and a flux of 1.1×1017nm−2 s−1 at 290°C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ′-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

Evaluation of residual stress near the weld overlay cladding by welding and post-weld heat treatment

Austenitic stainless steel is welded as a cladding on the inner surface of a reactor pressure vessel (RPV) made of low alloy steel. In order to assess the structural integrity of the RPV precisely, the residual stress distribution caused by weld-overlay cladding and post-weld heat treatment (PWHT) is evaluated. Since the cladding layer is very thin compared to the vessel wall, it is necessary to evaluate the residual stress distribution around the weld fusion line, which can be very steep. In this study, cladded specimens were fabricated using different welding methods. Residual stress measurements using both sectioning and deep hole drilling (DHD) methods were then performed to evaluate the residual stress distributions through the weld fusion line. Three-dimensional thermal-elastic-plastic-creep analyses based on the finite element method were also conducted to evaluate the residual stress caused by weld-overlay cladding and PWHT. It was shown that analytical results provided reasonable agreements on weld residual stress with experimental results. It was also clarified that the main cause of residual stress due to welding and PWHT was the difference of thermal expansion between weld and base metals.

Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the δ-ferrite phase but not in the austenitic phase. Thermal aging at 400°C for 10,000h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn=16:7:6 in the δ-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% δ-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M 23C 6 type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

原子炉圧力容器用確率論的破壊力学解析コードＰＡＳＣＡＬ２のクラッド部評価に関する機能改良

The probabilistic fracture mechanics analysis code PASCAL2 has been updated by Japan Atomic Energy Agency (JAEA) to support the validation of the codes and standards that provide the structural integrity assessment method of reactor pressure vessels during pressurized thermal shock. In this study, the evaluation methods for flaw initiation considering weld-overlay cladding and nondestructive inspection models and considering the existence of cladding have been improved. Using the improved PASCAL2 code, the effects of flaw initiation, the growth model for a cladding flaw, and the nondestructive inspection model on failure probability have been analyzed. The analysis verifies that the application of the improved flaw initiation model decreases the conditional probability of failure. The result also shows that only the maximum detection probability affects the failure probability among the parameters of the improved nondestructive inspection model.

Weld overlay cladding of high strength low alloy steel with austenitic stainless steel – Structure and properties

The present work aims at studying structure–property correlations in a weld overlay clad high strength low alloy steel with austenitic stainless steel of American Institute for Steel and Iron (AISI) 347 grade. Optical microscopy studies revealed that the interface between the two steels was nearly flat. The base plate had ferrite plus bainite microstructure adjacent to the interface and tempered bainite/martensite structure away from the interface. Grain coarsening and decarburization were observed near the interface. The stainless steel exhibited austenite dendritic structure. Tensile strength, notch-tensile strength and charpy impact energy of the base plate were found to be higher than those for the interface. The microhardness was observed to be maximum on the clad layer near interface. The shear bond strength of the weld overlay-interface was higher than the shear strength of the base plate. Fractography was carried out using scanning electron microscope on tensile, notch-tensile and shear bond test specimens of the interface as well as shear test specimens of the base plate. It revealed the presence of predominantly dimpled rupture. Charpy impact specimens of the interface failed in mixed mode while impact specimens of the base plate failed in ductile mode. Electron probe microanalysis across the bond interface indicated linear change in concentrations of Cr, Ni, Mn, Cu, Mo, Nb and Si between the levels appropriate to the clad layer and base metal.

Longer life with welded claddings

Alan Robinson of weld overlay cladding specialist Arc Energy Resources discusses the benefits of using the various techniques to protect the internal surfaces of pumps handling difficult media from corrosion. Here, he argues that this can be more cost-effective than using highly alloyed materials for the whole machine.

原子炉圧力容器肉盛溶接部における溶接及び溶接後熱処理による残留応力分布評価

Austenitic stainless steel is welded as a cladding on the inner surface of a reactor pressure vessel (RPV) made of low alloy steel. In order to assess the structural integrity of the RPV precisely, the residual stress distribution caused by weld-overlay cladding and post-weld heat treatment (PWHT) is evaluated. Since the cladding layer is very thin compared to vessel wall, it is necessary to evaluate the residual stress distribution around the weld fusion line can be very steep. In this study, cladded specimens were fabricated using different welding methods. Residual stress measurements using both sectioning and deep hole drilling (DHD) methods were then performed to evaluate the residual stress distributions through the weld fusion line. Three-dimensional thermal-elastic-plastic-creep analyses based on finite element method were also conducted to evaluate the residual stress caused by weld-overlay cladding and PWHT. It was shown that analytical results provided reasonable agreements on weld residual stress with experimental results. It was also clarified that the main cause of residual stress due to welding and PWHT was the difference of thermal expansion between weld and base metals.

肉盛溶接による熱影響部の材質変化を考慮した原子炉圧力容器の溶接残留応力評価

In order to evaluate residual stress distributions in a reactor pressure vessel (RPV) due to weld-overlay cladding and post-weld heat-treatment (PWHT), thermal-elastic-plastic-creep analysis considering the changes of material properties in heat-affected zone (HAZ) has been performed. By considering the material property changes in HAZ, stress distributions obtained from the analysis agree well with the experimental results measured from weld-overlay cladded plate. Applying the analysis method, residual stress distributions for a vessel model caused by weld-overlay cladding, PWHT, hydrostatic test, operational load and transient loads during pressurized thermal shock (PTS) events have been computed. Effects of the residual stresses on the structural integrity of RPVs have been evaluated by comparing the difference in stress distributions and the values of stress intensity factor for a postulated crack during PTS events.

Development of probabilistic fracture mechanics analysis codes for reactor pressure vessels and piping considering welding residual stress

Probabilistic fracture mechanics (PFM) analysis codes for reactor pressure vessels (RPVs) and piping, called as PASCAL (PFM Analysis of Structural Components in Aging LWRs) series, have been developed. The PASCAL2 (PASCAL version 2) evaluates the conditional probability of fracture of an RPV under transient conditions including pressurized thermal shock (PTS) considering neutron irradiation embrittlement of the vessels. Recent improvements to PASCAL2 are related to the treatment of weld–overlay cladding. The results using the improved code indicate that the residual stress by weld–overlay cladding affects the fracture probability to some extent. The PASCAL-SP (PASCAL – Stress corrosion cracking at welding joints for Piping) evaluates the probabilities of failures including leakage and breaks of safety-related piping complying with Japanese regulation and rules. Effects of welding residual stress distribution as well as inspection accuracy are focused in this study. Residual stress distributions have been determined by parametric FEM analyses and incorporated into the code.

Influence of Ti and C on the Solidification Microstructure of Fe-10Al-5Cr Alloys

The influence of Ti and C on the solidification microstructure of Fe-10Al-5Cr (all composition values in weight percent) alloys was examined with solidification modeling and a variety of experimental techniques. Several Fe-10Al-5Cr-Ti-C alloys were fabricated using the arc button melting process and characterized using quantitative image analysis and electron microscopy techniques. The experimental alloys exhibited primary ferrite dendrites with an interdendritic ferrite/TiC eutectic constituent, and the amount of eutectic was affected by the Ti and C concentrations. A liquidus projection and primary solidification paths were calculated for the Fe-10Al-5Cr-Ti-C system in order to estimate the amount of TiC that is expected to form during solidification. The range in the calculated amount of TiC-containing eutectic matched the experimentally measured values reasonably well. The ability to control the amount of TiC that forms during solidification of an Fe-10Al-5Cr-Ti-C-based alloy shows promise for developing corrosion-resistant weld overlay claddings with resistance to hydrogen cracking.

Effects of Weld-overlay Cladding on the Structural Integrity of Reactor Pressure Vessels during Pressurized Thermal Shock

Effects of Weld-overlay Cladding on the Structural Integrity of Reactor Pressure Vessels during Pressurized Thermal Shock

1322 肉盛溶接による熱影響部の材質変化を考慮した原子炉圧力容器の溶接残留応力評価(OS13.破壊力学とき裂の解析・き裂進展シミュレーション(6),オーガナイズドセッション)

原子炉圧力容器内面の肉盛溶接による溶接残留応力を評価するため、溶接に伴う熱影響部の材質変化を考慮した熱弾塑性クリープ解析を行った。熱影響部の材質変化を考慮することで、解析で得られた応力分布は平板型試験体から実測された応力分布に良く一致した。この解析手法を用い、原子炉圧力容器を模擬したモデルに対して肉盛溶接及び溶接後熱処理による残留応力解析を実施した。また、原子炉圧力容器に対する各種供用状態の熱弾塑性解析を行い、耐圧試験を含む通常運転荷重及び加圧熱衝撃時において溶接残留応力が構造健全性に及ぼす影響を応力拡大係数の観点から評価した。

Weld overlay cladding – the solution to pump corrosion?

Alan Brown, of Arc Energy Resources discusses the benefits of weld overlay cladding to protect the internal surfaces of oil and gas pumping equipment from corrosion, and reviews the options available when using corrosion resistant alloys.