Nozzle Dissimilar Metal Welds Research Articles

Study on Micro-Structure and Tensile Mechanical Properties of Dissimilar Metal Weld Joint Connecting Steam Generator Nozzle and Safe-End

The safe-end of a steam generator (SG) nozzle dissimilar metal weld (DMW) for pressurized water reactors (PWRs) is the weakest point of failure which is crucial for the safe operation of a nuclear power station. Related to materials micro-structures, a uniaxial stress–strain relationship is the basic input parameter for nuclear power plant design, safety evaluation, and life management. In this paper, the micro-structure and tensile mechanical properties of a DMW of a European pressurized water reactor (EPR) were studied. Vickers hardness tester, optical microscope, and electron back scatter diffraction were used to analyze the micro-structure of the DMW joint. In addition, the residual strain of the DMW joint base material, heat-affected zone, weld metal, and fusion boundary region were studied. Based on digital image correlation (DIC) technology, tensile mechanical properties of the DMW joint were obtained. The results show that an accurate tensile stress–strain relationship of dissimilar metal welded joints can be obtained by using the DIC technique, the weld is the relatively weak link, and the residual strain is concentrated in the heat-affected zone. This study provides valuable engineering information regarding nuclear power plant design, in-service performance testing, and structural analysis and evaluation of welds containing defects.

Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

There have been incidents recently where stress corrosion cracking (SCC) observed in the dissimilar metal weld (DMW) joints connecting the reactor pressure vessel (RPV) nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR). Tensile residual stress (RS), in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM), an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

Weld Residual Stress Analysis and the Effects of Structural Overlay on Various Nuclear Power Plant Nozzles

Welding is a commonly used and one of the most important material-joining processes in industry. The incidences of defects had been located by ultrasonic testing in various pressurizer nozzle dissimilar metal welds (DMW) at nuclear power plants. In order to evaluate the crack propagation, it is required to calculate the stress distribution including weld residual stress and operational stress through the wall thickness in the weld region. The analysis procedure in this paper included not only the pass-by-pass welding steps but also other essential fabrication steps of surge, safety/relief, and spray nozzles. In this paper, detailed welding simulation analyses have been conducted to predict the magnitude of these stresses in the weld material. To prevent primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR) on susceptible welded pipes with dissimilar metal welds, the weld overlay process has been applied to repair nuclear reactor pipe joints in plants. The objectives of such repairs are to induce compressive axial residual stresses on the pipe inside surface, as well as increase the pipe thickness with a weld material that is not susceptible to stress corrosion cracking. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe joints with weld overlay repairs. The finite element results in this paper showed that, after deposition of the DMW nozzle and stainless steel welds, tensile weld residual stresses still exist at regions of the DMW through the thickness. This tensile weld residual stress region was significantly reduced after welding the overlay. The overlay weld also provides a more uniform and large compressive region through the thickness, which has a beneficial effect on the structural integrity of the DMW in the plant.

Stress Corrosion Crack Shape Development Using AFEA

Typical ASME Section XI subcritical cracking analyses assume an idealized flaw shape driven by stress intensity factors developed for semi-elliptical shaped flaws. Recent advanced finite element analyses (AFEA) conducted by both the United States Nuclear Regulatory Commission (U.S.NRC) and the nuclear industry for long circumferential indications found in the pressurizer nozzle dissimilar metal welds at the Wolf Creek power plant suggest that the semi-elliptical flaw assumption may be overly conservative in some cases. The AFEA methodology that was developed allowed the progression of a planar flaw subjected to typical stress corrosion cracking (SCC)-type growth laws by calculating stress intensity factors at every nodal point along the crack front, and incrementally advancing the crack front in a more natural manner. Typically, crack growth analyses increment the semi-elliptical flaw by considering only the stress intensity factor at the deepest and surface locations along the crack front, while keeping the flaw shape semi-elliptical. In this paper, a brief background to the AFEA methodology and the analyses conducted in the Wolf Creek effort will be discussed. In addition, the predicted behavior of surface cracks under normal operating conditions (plus welding residual stress) using AFEA will be investigated and compared with the semi-elliptical assumption. Conclusions on the observation of when semi-elliptical flaw assumptions are appropriate will be made. These observations will add insight into the conservatism of using an idealized flaw shape assumption.