Simulation of Triaxial Residual Stress Mapping for a Hollow Cylinder

Author(s):  
Mitchell D. Olson ◽  
Wilson Wong ◽  
Michael R. Hill

This paper describes a novel method to determine a two-dimensional map of the triaxial residual stress on a radial-axial plane of interest in a hollow cylindrical body. With the description in hand, we present a simulation to validate the steps of the method. The simulation subject is a welded cylindrical nozzle typical of a nuclear power pressurized water reactor pressurizer; in the weld region, the nozzle inner diameter is roughly 132 mm (5.2 inch) and the wall thickness is roughly 35 mm (1.4 inch). The pressure vessel side of the nozzle is carbon steel (with a thin stainless steel lining), the piping side is austenitic stainless steel, and between the two are weld and buttering deposits of nickel alloy. Weld residual stresses in such nozzles have important effects on crack growth rates in fatigue and stress corrosion cracking, therefore measurements of weld residual stress can help provide inputs for managing aging reactor fleets. Nuclear power plant welds often have large and complex geometry, which has made residual stress measurements difficult, and this work provides a proof of concept for a new experimental technique for measurements on welded nozzles.

Author(s):  
F. W. Brust ◽  
Tao Zhang ◽  
Do-Jun Shim ◽  
Sureshkumar Kalyanam ◽  
Gery Wilkowski ◽  
...  

Flaw indications have been found in some dissimilar metal nozzle to stainless steel piping welds in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe using Alloy 182/82 weld metal. Due to an unexpected aging issue with the weld metal, the weld becomes susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually >300C) and the water chemistry in the PWR is typical of operating plants. This paper represents one of a series of papers which examine the propensity for cracking in a particular operating PWR in the UK. This paper represents an examination of the weld residual stress distributions which occur in four different size nozzles in the plant. Companion papers in this conference examine crack growth and PWSCC mitigation efforts related to this plant. British Energy (BE) has developed a work program to assess the possible impact of PWSCC on dissimilar metal welds in the primary circuit of the Sizewell ‘B’ pressurized water reactor. This effort has included the design and manufacture of representative PWR safety/relief valve nozzle welds both with and without a full structural weld overlay, multiple residual stress measurements on both mock-ups using the deep hole and incremental deep hole methods, and a number of finite element weld residual stress simulations of both the mock-ups and equivalent plant welds. This work is summarized in companion papers [1–3]. Here, the detailed weld residual stress predictions for these nozzles are summarized. The weld residual stresses in a PWR spray nozzle, safety/relief nozzle, surge nozzle, and finally a steam generator hot-leg nozzle are predicted here using an axis-symmetric computational weld solution process. The residual stresses are documented and these feed into a natural crack growth analysis provided in a companion PVP 2010-25162 paper [1]. The solutions are made using several different constitutive models: kinematic hardening, isotropic hardening, and a mixed hardening model. Discussion will be provided as to the appropriateness of the constitutive model for multi-pass DM weld modeling. In addition, the effect of including or neglecting the post-weld heat treatment process, which typically occurs after the buttering process in a DM weld, is presented. During operation the DM welds in a PWR experience temperatures in excess of 300°C. The coefficient of thermal expansion (CTE) mismatch between the three materials, particularly the higher CTE in the stainless steel, affects the stresses at operating temperature. The K-weld geometry used in the steam generator nozzles in this plant combines with CTE mis-match effects to result in service stresses somewhat different from V-weld groove cases.


2019 ◽  
Vol 822 ◽  
pp. 53-59 ◽  
Author(s):  
Anton Sergeevich Tsvetkov ◽  
Irina Vladimirovna Teplukhina

Currently, a new generation of pressurized water reactor for nuclear power plants with an extended service life (up to 60 years) and a guarantee of their complete safety are being designed in Russia. Analysis of the reactor internal elements performance showed, that designed service life cannot be guaranteed if the reactor’s internal parts would be made from currently used stainless steel (18-10 alloy type). Instead of the used steel, to ensure operability, new austenitic stainless steel (16-25 alloy type), with increased resistance to radiation swelling, is being developed for production of forged ring blanks for core baffle. The use of new steel requires revision of the existing metallurgical production technology stages. Therefore in this paper diffusion experiment was carried out to determine the required duration of homogenization. The results are presented in terms of different duration of the high-temperature exposure effect on the liquation heterogeneity equalization. The relation between duration of homogenization and microhardness is also shown.


Author(s):  
Francis H. Ku ◽  
Shu (Stan) S. Tang

Finite element weld residual stress analyses are performed to investigate the similarities and differences between two-dimensional (2-D) and three-dimensional (3-D) finite element analyses on weld residual stress predictions of the NRC Phase II Mockup. The Mockup resembles a typical pressurized water reactor (PWR) surge nozzle of 14″ in diameter which includes a dissimilar metal weld (DMW) connecting the safe end and a stainless steel weld (SSW) connecting the surge line piping. The 2-D analysis employs axisymmetric modeling approach, while the 3-D analysis utilizes moving heat source approximation techniques. The results demonstrate the variations in residual stresses among the weld bead start and stop locations. Comparing the 2-D and 3-D residual results against experiment measurements also reveal the limitations inherent to the 2-D analysis, while the 3-D analysis can produce results that are of closer match to experimental measurements.


Author(s):  
Frederick W. Brust ◽  
E. Punch ◽  
D. J. Shim ◽  
David Rudland ◽  
Howard Rathbun

Flaw indications have been found in some dissimilar metal (DM) nozzle to stainless steel piping welds and reactor pressure vessel heads (RPVH) in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe) using Alloy 182/82 weld metal. The welds may become susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually greater than 300°C) and the water chemistry in the PWR is typical of operating plants. The weld residual stresses (WRS) induced by the welds are a main driver of PWSCC. The purpose of this paper is to determine the weld residual stresses in a double-vee groove welded nozzle and then to model the natural crack growth in the weld. The double vee groove geometry has not been modeled much to date especially in such a large nozzle. This leads to a rather unique weld residual stress pattern which changes as the throat of the double vee is approached. Axial crack growth is modeled using a natural crack growth procedure. This was challenging since the crack shape necked down in the region where the tips of the vee grooves meet making the mesh development during this process challenging. This analysis provides information regarding crack growth evolution versus time. In addition, some comments regarding idealized growth are presented.


Author(s):  
Michael R. Hill ◽  
Mitchell D. Olson

This paper describes a sequence of residual stress measurements made to determine a two-dimensional map of biaxial residual stress in a dissimilar metal welded nozzle typical of a nuclear pressurized water reactor (PWR). The present experimental work follows on the numerical analysis reported earlier, in PVP2012-78885. The measurement subject is a cylindrical nozzle, removed from a PWR pressure vessel, having a nickel alloy weld joining a stainless steel safe end to a low-alloy steel vessel. Biaxial residual stress was determined in a series of experimental steps using strain gage measurements, the contour method, and slitting. Confirmatory measurements were also performed (including digital image correlation and neutron diffraction). The paper includes descriptions of the experimental steps, data reduction, and residual stress results, along with a comparison between measurements and output from a weld simulation. The measured hoop stress in the weld region is tensile near the OD (300 MPa) and compressive near the ID (−400 MPa); the measured axial stress is tensile near the OD (150 MPa) and compressive near the ID (−150 MPa).


Author(s):  
Ying Hong ◽  
Xuesheng Wang ◽  
Yan Wang ◽  
Zhao Zhang ◽  
Yong Han

Stainless steel 304 L tubes are commonly used in the fabrication of heat exchangers for nuclear power stations. The stress corrosion cracking (SCC) of 304 L tubes in hydraulically expanded tube-to-tubesheet joints is the main reason for the failure of heat exchangers. In this study, 304 L hydraulically expanded joint specimens were prepared and the residual stresses of a tube were evaluated with both an experimental method and the finite element method (FEM). The residual stresses in the outer and inner surfaces of the tube were measured by strain gauges. The expanding and unloading processes of the tube-to-tubesheet joints were simulated by the FEM. Furthermore, an SCC test was carried out to verify the results of the experimental measurement and the FEM. There was good agreement between the FEM and the experimental results. The distribution of the residual stress of the tube in the expanded joint was revealed by the FEM. The effects of the expansion pressure, initial tube-to-hole clearance, and yield strength of the tube on the residual stress in the transition zone that lay between the expanded and unexpanded region of the tube were investigated. The results showed that the residual stress of the expanded joint reached the maximum value when the initial clearance was eliminated. The residual stress level decreased with the decrease of the initial tube-to-hole clearance and yield strength. Finally, an effective method that would reduce the residual stress without losing tightness was proposed.


Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.


Author(s):  
Michael L. Benson ◽  
Patrick A. C. Raynaud ◽  
Frederick W. Brust

Residual stress prediction contributes to nuclear safety by enabling engineering estimates of component service lifetimes. Subcritical crack growth mechanisms, in particular, require residual stress assumptions in order to accurately model the degradation phenomena. In many cases encountered in nuclear power plant operations, the component geometry permits two-dimensional (i.e., axisymmetric) modeling. Two recent examples, however, required three-dimensional modeling for a complete understanding of the weld residual stress distribution in the component. This paper describes three-dimensional weld residual stress modeling for two cases: (1) branch connection welds off reactor coolant loop piping and (2) a mockup to demonstrate the effectiveness of the excavate and weld repair process.


1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.


Author(s):  
Tao Zhang ◽  
F. W. Brust ◽  
Gery Wilkowski

Weld residual stresses in nuclear power plant can lead to cracking concerns caused by stress corrosion. These are large diameter thick wall pipe and nozzles. Many factors can lead to the development of the weld residual stresses and the distributions of the stress through the wall thickness can vary markedly. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe and nozzle joints with welds. This paper represents an examination of the weld residual stress distributions which occur in various different size nozzles. The detailed weld residual stress predictions for these nozzles are summarized. Many such weld residual stress solutions have been developed by the authors in the last five years. These distributions will be categorized and organized in this paper and general trends for the causes of the distributions will be established. The residual stress field can therefore feed into a crack growth analysis. The solutions are made using several different constitutive models such as kinematic hardening, isotropic hardening, and mixed hardening model. Necessary fabrication procedures such as repair, overlay and post weld heat treatment are also considered. Some general discussions and comments will conclude the paper.


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