Crack Growth Behavior of Nickel Alloy Welds in a PWR Environment

2007 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Crack Growth ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Crack Growth Behavior ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Ni Alloy ◽  
The Individual

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

scholarly journals Crack Growth Rates of Nickel Alloy Welds in a PWR Environment

2006 ◽  
Author(s):  
Bogdan Alexandreanu ◽  
Omesh K. Chopra ◽  
William J. Shack
Keyword(s):  
Weld Metal ◽  
Crack Growth ◽  
Growth Rates ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Average Growth ◽  
Deep Groove ◽  
Ni Alloy

A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This paper presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320°C. Crack growth tests were conducted on 1–T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The fatigue CGRs of Alloy 182 in the PWR environment are a factor of ≈5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking CGRs for Alloy 182 are close to the average growth rates of Alloy 600 in the PWR environment. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

Fatigue Strain–Life Behavior of Austenitic Stainless Steels in Pressurized Water Reactor Environments

2015 ◽  
Author(s):  
Paul Wilhelm ◽  
Paul Steinmann ◽  
Jürgen Rudolph
Keyword(s):  
Fatigue Life ◽  
Stainless Steels ◽  
National Laboratory ◽  
Reference Conditions ◽  
Argonne National Laboratory ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

A statistical model for austenitic stainless steels for predicting the effect of pressurized water reactor environments on fatigue life for a range of temperatures and strain rates is developed based on analysis of available material data. The compiled fatigue curve data include not only results from America (Keller (1971), Conway (1975), Hale (1977), and Argonne National Laboratory (1999–2005)), but also from Europe (Solin (2006), Le Duff (2008–2010), De Baglion (2011, 2012), Huin (2013) …) and Japan (Kanasaki (1997)). Only fatigue data from polished specimens of wrought material tested under strain control were considered; hollow specimens were not treated herein. The fatigue life correction factor used in this paper was defined as the ratio of life in water at 300 °C (reference conditions) to that in water at service conditions. The model is recommended for predicting fatigue lives that are 103–105 cycles.

Stress Corrosion Cracking Behavior of Alloy 182 Weld in Pressurized Water Reactor Primary Water Environment at 325ºC

CORROSION ◽  
2011 ◽  
Vol 67 (8) ◽  
pp. 085004-1-085004-9 ◽  
Author(s):  
L.I.L. Lima ◽  
M.M.A.M. Schvartzman ◽  
C.A. Figueiredo ◽  
A.Q. Bracarense
Keyword(s):  
Stress Corrosion ◽  
Oxide Layer ◽  
Stress Corrosion Cracking ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Primary Water

Abstract The weld used to connect two different metals is known as a dissimilar metal weld (DMW). In nuclear power plants, this weld is used to join stainless steel to low-alloy steel components in the nuclear pressurized water reactor (PWR). The most common weld metal is Alloy 182 (UNS W86182). Originally selected for its high corrosion resistance, it exhibited, after a long operation period, susceptibility to stress corrosion cracking (SCC) in PWR. The goal of this work was to study the electrochemical corrosion behavior and SCC susceptibility of Alloy 182 weld in PWR primary water containing 25 cm3 and 50 cm3 H2/kg H2O at standard temperature and pressure (STP). For this purpose, slow strain rate tensile (SSRT) tests and potentiodynamic polarization measurements were carried out. Scanning electron microscopy (SEM) with energy-dispersive spectrometry (EDS) was used to evaluate fracture morphology and determine the oxide layer chemical composition and morphology. The results indicated that at 325°C Alloy 182 weld is more susceptible to SCC at 25 cm3 (STP) H2/kg H2O and the increase of dissolved hydrogen decreased the crystal size of the oxide layer.

Effect of Uneven Crack Front on Crack Tip Mechanics and the Implication to Stress Corrosion Crack Growth

2009 ◽  
Author(s):  
He Xue ◽  
Zhanpeng Lu ◽  
Hiroyoshi Murakami ◽  
Tetsuo Shoji
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Crack Front ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Nickel Base

Uneven crack fronts have been observed in laboratory stress corrosion cracking tests. For example, cracking fronts of nickel-base alloys tested in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments could exhibit uneven crack front. Analyzing the effect of an uneven crack front on further crack growth is important for quantification of crack growth. Finite-Element analysis shows that the local KI distribution can be significantly affected by the shape and size of the uneven crack front. Stress intensity factor at the locally extended crack front can be significantly reduced. Since generally there is a nonlinear CGR versus KI relationship, it is expected that crack growth rate at the locally extended crack front can be significantly different from those in the neighboring areas. There could be several patterns for the growth of an uneven crack front. For example, once initiated, the crack growth rate in areas other than the locally protruded front would become higher and then the whole crack front would tend to become uniform. On the other hand, if the crack growth in other areas is still low, there is a possibility that the crack growth rate at the front tip would slow down.

Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China

2010 ◽  
Author(s):  
Yan Wang ◽  
Xie Heng
Keyword(s):  
Uncertainty Analysis ◽  
System Analysis ◽  
High Efficiency ◽  
Current System ◽  
National Laboratory ◽  
Pressurized Water Reactor ◽  
Economic Competition ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

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