Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

2011 ◽  
Author(s):  
Frederick W. Brust ◽  
R. E. Kurth ◽  
D. J. Shim ◽  
David Rudland
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Welding Process ◽  
Corrosion Cracking ◽  
Uncertainty Sources

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Stress Corrosion Cracking and Corrosion Fatigue Analysis of Nuclear Steam Turbine Rotor

2014 ◽  
Author(s):  
Gang Chen ◽  
Puning Jiang ◽  
Xingzhu Ye ◽  
Junhui Zhang ◽  
Yifeng Hu ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Turbine ◽  
Corrosion Fatigue ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Cracking ◽  
Corrosion Cracking ◽  
Major Influence ◽  
Steam Turbines

Although stress corrosion cracking (SCC) and corrosion fatigue cracking can occur in many locations of nuclear steam turbines, most of them initiate at low pressure disc rim, rotor groove and keyway of the shrunk-on disc. For nuclear steam turbine components, long life endurance and high availability are very important factors in the operation. Usually nuclear power plants operating more than sixty years are susceptible to this failure mechanism. If SCC or corrosion fatigue happens, especially in rotor groove or keyway, it has a major influence on nuclear steam turbine life. In this paper, established methods for the SCC and corrosion fatigue-controlled life prediction of steam turbine components were applied to evaluating a new shrunk-on disc that had suffered local keyway surface damage during manufacture and loss of residual compressive stress.

Application of Failure Event Data to Benchmark Probabilistic Fracture Mechanics Computer Codes

2007 ◽  
Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
B. O. Y. Lydell ◽  
D. L. Rudland ◽  
G. M. Wilkowski
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Operating Experience ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Probabilistic Fracture Mechanics ◽  
Computer Codes

This paper describes an application of data on cracking, leak and rupture events from nuclear power plant operating experience to estimate failure frequencies for piping components that had been previously evaluated using the PROLOCA and PRAISE probabilistic fracture mechanics (PFM) computer codes. The calculations had addressed the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking and fatigue for materials and operating conditions that were known to have failed components. The first objective was to benchmark the calculations against field experience. A second objective was a review of uncertainties in the treatments of the data from observed failures and in the structural mechanics models. The database PIPExp-2006 was applied to estimate failure frequencies. Because the number of reported failure events was small, there were also statistical uncertainties in the estimates of frequencies. Comparisons of predicted and observed failure frequencies showed that PFM codes correctly predicted relatively high failure probabilities for components that had experienced field failures. However, the predicted frequencies tended to be significantly greater than those estimated from plant operating experience. A review of the PFM models and inputs to the models showed that uncertainties in the calculations were sufficiently large to explain the differences between the predicted and observed failure frequencies.

scholarly journals Fitness-for-Service Assessment for a Radioactive Waste Tank That Contains Stress Corrosion Cracks

2009 ◽  
Author(s):  
Bruce J. Wiersma ◽  
James B. Elder ◽  
Rodney W. VandeKamp ◽  
Charles A. McKeel
Keyword(s):  
Residual Stresses ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Degradation Mechanism ◽  
Service Condition ◽  
Corrosion Cracking ◽  
The Future ◽  
Perpendicular Crack ◽  
The Impact

Radioactive wastes are confined in 49 underground storage tanks at the Savannah River Site. The tanks are examined by ultrasonic (UT) methods for thinning, pitting, and stress corrosion cracking in order to assess fitness-for-service. During an inspection in 2002, ten cracks were identified on one of the tanks. Given the location of the cracks (i.e., adjacent to welds, weld attachments, and weld repairs), fabrication details (e.g., this tank was not stress-relieved), and the service history the degradation mechanism was stress corrosion cracking. Crack instability calculations utilizing API-579 guidance were performed to show that the combination of expected future service condition hydrostatic and tensile weld residual stresses do not drive any of the identified cracks to instability. The cracks were re-inspected in 2007 to determine if crack growth had occurred. During this re-examination, one indication that was initially reported as a “possible perpendicular crack <25% through wall” in 2002, was clearly shown not to be a crack. Additionally, examination of a new area immediately adjacent to other cracks along a vertical weld revealed three new cracks. It is not known when these new cracks formed as they could very well have been present in 2002 as well. Therefore, a total of twelve cracks were evaluated during the re-examination. Comparison of the crack lengths measured in 2002 and 2007 revealed that crack growth had occurred in four of the nine previously measured cracks. The crack length extension ranged from 0.25 to 1.8 inches. However, in all cases the cracks still remained within the tensile weld residual stress zone (i.e., within two to three inches of the weld). The impact of the cracks that grew on the future service of Tank 15 was reassessed. API-579 crack instability calculations were again performed based on expected future service conditions and trended crack growth rates for the future tank service cycle. The analysis showed that the combined hydrostatic and tensile weld residual stresses do not drive the identified cracks to instability. This tank expected to be decommissioned in the near future. However, if these plans are delayed, it was recommended that a third examination of selected cracks in the tank be performed in 2014.

scholarly journals Stress Corrosion Cracking Behavior of TP 439 and 690 TT under Film-Forming Amine Environment

Scanning ◽  
2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Lu Jundong ◽  
Jiang Xiaobin ◽  
Sun Ke ◽  
Liu Bin ◽  
Li Xinmin ◽  
...  
Keyword(s):  
Stainless Steel ◽  
High Temperature ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Cracking Behavior ◽  
Film Forming ◽  
Metal Materials

Film-forming amines have been widely used in thermal power plants for maintenance after shutdown, and there are more and more applications and researches in nuclear power secondary circuits for this purpose. However, in the direction of stress corrosion cracking, there is not much research on the influence of film-forming amines on metal materials. This article uses the high temperature slow strain rate test (SSRT) method to evaluate the influence of a commercial film-forming amine on the stress corrosion cracking behavior of two conventional island materials for PWR nuclear power plants. These two metal materials are the heat exchange tube materials of the high-pressure heater and steam generator in the high-temperature operation area of the secondary circuit of a nuclear power plant: TP 439 stainless steel and 690 TT alloy. The test analyzed the mechanical properties and fracture morphology. The test results show that in the test concentration range (<5 mg/kg), the film-forming amine will not affect the SCC of TP 439 stainless steel and 690 TT alloy under the condition of slow strain rate. The behavior has a significant impact. In practical applications, the general dosage of film-forming amine is 1-2 mg/kg. This data is lower than the film-forming amine concentration used in the experiment. Therefore, there is no need to worry about the obvious impact on the SCC behavior of TP 439 stainless steel and 690 TT alloy.

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