Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds

2008 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Aladar A. Csontos ◽  
Timothy R. Lupold ◽  
Chia-Fu Sheng
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Primary Water ◽  
Water Reactors ◽  
The Impact ◽  
Growth Analyses

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.

Through-Thickness Welding Residual Stress Profile in Dissimilar Metal Nozzle Butt Weld

2010 ◽  
Author(s):  
Tae-Kwang Song ◽  
Ji-Soo Kim ◽  
Chang-Young Oh ◽  
Hong-Yeol Bae ◽  
Jun-Young Jeon ◽  
...  
Keyword(s):  
Residual Stress ◽  
Welding Residual Stress ◽  
Stress Profile ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Residual Stress Profile ◽  
Dissimilar Metal ◽  
Stress Profiles ◽  
Water Reactors

This paper provides the through-thickness welding residual stress profile in dissimilar metal nozzle butt welds of pressurized water reactors. For systematic investigations of the effects of geometric variables, i.e. the thickness and the radius of the nozzle and the length of the safe end, on welding residual stresses, idealized shape of nozzle is proposed and elastic-plastic thermo-mechanical finite element analyses are conducted. Through-wall welding residual stress profiles for dissimilar metal nozzle butt welds are proposed, which take a modified form of existing welding residual stress profiles developed for austenitic pipe butt weld in R6 code.

Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

PWSCC Crack Growth Modeling Approaches

2011 ◽  
Author(s):  
F. W. Brust ◽  
D.-J. Shim ◽  
G. Wilkowski ◽  
D. Rudland
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Intensity ◽  
Residual Stresses ◽  
Crack Growth ◽  
Corrosion Cracking ◽  
Growth Time ◽  
Pressurized Water ◽  
Primary Water ◽  
Water Reactors

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 >300C) 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. Several mechanical mitigation methods to control PWSCC have been developed for use on a nozzle welds in nuclear PWR plants. These methods consist of applying a weld overlay repair (WOR), using a method called mechanical stress improvement process (MSIP), and applying an inlay to the nozzle ID. The purpose of a mitigation method is to reduce the probability that PWSCC will occur in the nozzle joint. The key to assessing the effectiveness of mitigation is to determine the crack growth time to leak with and without the mitigation. Indeed, for WOR and MSIP, the weld residual stresses are often reduced after application while for inlay they are actually increased. However, all approaches reduce crack growth rates if applied properly. Procedures for modeling PWSCC growth tend to vary between organizations performing the analyses. Currently, the prediction of PWSCC crack growth is based on the stress intensity factors at the crack tips. Several methods for evaluating the stress intensity factor for modeling the crack growth through these WRS fields are possible, including using analytical, natural crack growth using finite element methods, and using the finite element alternating method. This paper will summarize the methods used, critique the procedures, and provide some examples for crack growth with and without mitigation. Suggestions for modeling such growth will be provided.

Initiation of PWSCC in Nickel Base Alloy 182

2013 ◽  
Author(s):  
Mickaël Wehbi ◽  
Jérôme Crépin ◽  
Thierry Couvant ◽  
Cécilie Duhamel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Alloy ◽  
Nickel Base Alloy ◽  
Pressurized Water ◽  
Material Environment ◽  
Primary Water ◽  
Water Reactors ◽  
Nickel Base ◽  
Actual Initiation

To date, welded nickel base alloy 182 used in Pressurized Water Reactors (PWRs) components have shown a higher susceptibility to Primary Water Stress Corrosion Cracking (PWSCC) during laboratory tests than in power plants. However, an increasing number of cracks reported in American, Swedish and Japanese nuclear power plants on Alloy 182 enlighten the need for a predictive initiation model of PWSCC. Initiation of PWSCC involves several factors such as material, environment and loading history, interacting with each other. Building such a model first requires to focus on these parameters separately, in order to have a better understanding of the involved mechanisms at a local scale in crack initiation. This study focuses on the correlation between EBSD/strain field results to improve the accuracy of the actual initiation model [1] involving local parameters.

Crack Growth Analysis in Alloy 82/182 Butt Weld by Using the Finite Element Alternating Method

2009 ◽  
Author(s):  
Maan-Won Kim ◽  
Young-Jong Kim ◽  
Byoung Chul Kim
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Growth Analysis ◽  
Welding Residual Stress ◽  
Analysis Model ◽  
Element Analysis ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Alternating Method

Primary water stress corrosion cracking (PWSCC) of Alloy 82/182 butt welds has been a concern for pressurized water reactor (PWR) plants worldwide for the past decade. A lot of works have been performed to calculate exact welding residual stresses and PWSCC growth rate for Alloy 82/182 butt welds. The PWSCC growth analysis of Alloy 82/182 butt welds has been performed by using the Raju-Newman type solutions for the crack tip stress intensity factors (SIFs). In this study, a finite element alternating method (FEAM) was used to calculate the SIFs and crack propagation in Alloy 82/182 butt weld. The FEAM is consisted of two solution parts: an exact theoretical SIF solution for a crack embedded in infinite body and a simple finite element analysis model with coarse mesh for finite body. In this study, the theoretical SIFs were derived by using a dislocation density function. Finite element (FE) analysis was performed to obtain the welding residual stress distribution for a nozzle with Alloy 82/182 butt weld and PWSCC growth analysis was performed by using the FEAM with PWSCC growth model described in MRP reports under welding residual stress along the nozzle thickness.

Verification Analyses for Newly Developed Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Do-Jun Shim ◽  
Gery Wilkowski
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
3D Finite Element ◽  
Primary Water ◽  
Axial Crack ◽  
Analysis System

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Sign in / Sign up

Export Citation Format

Share Document