Probabilistic ageing and risk analysis tools for nuclear piping

Influence of crystallographic texture on susceptibility to stress corrosion cracking mechanism of API 5L X52 steel for sour service

Synchrotron-based holotomography and X-ray fluorescence study on the stress corrosion cracking behavior of the peak-aged 7075 aluminum alloy

The selective dissolution of ferrite phase from the pearlite was studied in fuel-grade ethanol (FGE) to understand how it affects the stress corrosion cracking (SCC) mechanism of carbon steel in FGE. It was shown that microgalvanic coupling occurs between ferrite and cementite phases of the pearlite, leading to localized corrosion, which affects the SCC mechanism. The intergranular SCC mechanism stops at the pearlite, and the selective dissolution promotes the transgranular SCC mechanism. Cathodic polarization curves were measured for pure iron and cementite exposed to various FGE conditions. According to the results, cementite phase is, in most cases, a more favorable cathode in FGE.

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.

In recent years, the compound beyond-design-basis accident (BDBA), which combines earthquake, tsunami, or some other severe events to impact a nuclear power plant (NPP), has received more attention. After the Fukushima nuclear disaster, the licensee of NPPs in Taiwan established the ultimate response guideline (URG) that instructs operators to perform reactor depressurization, low-pressure water injection, and containment venting to prevent core meltdown and hydrogen explosion once long-term loss-of-power and water-supply events occur. In this paper, we employed the probabilistic fracture mechanics (PFM) method to evaluate the structural integrity of boiling water reactor (BWR) pressure vessels under URG operation. At first, models of the beltline shell welds for BWR vessels associated with the Pressure Vessel Research Users Facility-Exponential flaw distribution were built for the PFM Fracture Analysis of Vessels–Oak Ridge (FAVOR) code. Then, the thermal-hydraulic data of URG transients for Taiwan domestic BWRs were imposed as the loading conditions. The analysis results demonstrate that performing URG operation will not cause significant fracture probability even at extreme embrittlement conditions. If long-term station blackout occurs due to a compound BDBA, the URG procedures can prevent core damage and hydrogen explosion, while maintaining the structural integrity of the reactor pressure vessels.

This paper describes Probabilistic Fracture Mechanics (PFM) analyses for quantitative evaluation of the likelihood of Leak-Before-Break (LBB) of Light Water Reactor's (LWR's) piping. The PFM analyses in general assume probabilistic distributions of initial crack size, applied stress cycles, crack growth laws, fracture criteria, leakage detection capability, defect inspection capability and so on. Referring to the deterministic procedure for LBB evaluation, most appropriate PFM models and data for LBB evaluation are discussed. Here the LBB index is newly proposed in order to quantitatively evaluate the likelihood of LBB. Through intensive sensitivity analyses, it is clarified that the LBB is more likely to occur for larger diameter pipe; the performance of leakage detection significantly affects the LBB likelihood; the LBB likelihood increases with plant's aging even conservatively assuming leak detection capability; the R6 method (Category 1, Option 1) for fracture criterion gives very conservative results; and In-Service Inspection (ISI) reduces the increase rate of failure probability than the failure probability itself.

Experimental study on the floor responses of a base-isolated frame structure via shaking table tests

An experimental framework for simulating stress corrosion cracking in cable bolts

Comparative study on the stress corrosion cracking of a new Ni-advanced high strength steel prepared by TMCP, direct quenching, and quenching & tempering

This paper reviews the relevant literature and covers the main aspects of the environmentally assisted cracking of Al-Mg-Si-(Cu) alloys. Apart from a brief overview of the major microstructural and mechanical properties, it presents research results on the corrosion sensitivity and stress corrosion susceptibility of Al-Mg-Si alloys. Possible mechanisms of stress corrosion cracking and corrosion fatigue in aluminum alloys, such as anodic dissolution and/or interaction with hydrogen, are considered. A number of factors, including atmospheric or solution conditions, applied stress, and material properties, can affect these mechanisms, leading to environmentally assisted cracking. Specific attention is given to Al-Mg-Si alloys with copper, which may increase the sensitivity to intergranular corrosion. The susceptibility to both intergranular corrosion and stress corrosion cracking of Cu-containing Al-Mg-Si alloys is mostly associated with a very thin layer (segregation) of Cu on the grain boundaries. However, the effect of Cu on the corrosion fatigue and fatigue crack growth rate of Al-Mg-Si alloys has received limited attention in the literature. At the current state of the research, it has not yet been holistically assessed, although a few studies have shown that a certain content of copper can improve the resistance of aluminum alloys to the environment with regard to corrosion fatigue. Furthermore, considerations of the synergistic actions of various factors remain essential for further studying environmentally assisted cracking phenomena in aluminum alloys.

Stress corrosion cracking in light water reactor is one of the most important factors threatening the safe operation of nuclear power plants. Due to the severity, generality and various safety and economic problems caused by this phenomenon, it is necessary to establish a model for predicting the stress corrosion cracking growth rates. This paper provides an overview of three main methods for predicting stress corrosion cracking growth rates in recent decades, i.e., empirical, deterministic and calculation methods, which are introduced in detail. Empirical models describe classical statistical analysis and emerging artificial neural network method, both of which are based on a large number of experimental test data mining. They are convenient and relatively accurate in predicting, but require extensive, time-consuming and expensive tests for different service environments. Deterministic models aim to establish a theoretical relationship between crack growth rate and various influencing parameters by studying the stress corrosion cracking mechanism. Many scholars have proposed different mechanisms to scientifically explain the stress corrosion cracking phenomenon and propose corresponding crack growth rate models. Calculation models reveal the mechanism of crack initiation and propagation in different layers of materials by means of finite element method based on fracture mechanics and multiscale method based on quantum mechanics. They provide new idea for future research on stress corrosion cracking and bridge the quantitative mechanism or model, but no specific stress corrosion cracking growth rate model is formed. The article concludes with the prospect, aim and direction for stress corrosion cracking mechanism and prediction model.

Structural integrity assessments for cracked nuclear components are currently performed on the basis of deterministic fracture mechanics in accordance with the Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers. On the other hand, probabilistic fracture mechanics (PFM) is expected as a rational method for a structural integrity assessment because it can account for the uncertainties and scatters of various influencing factors and can evaluate quantitative values such as the failure probabilities of the components as the solutions. In the Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed in order to evaluate the failure probability of nuclear pipe by taking into account aging degradation mechanisms such as inter-granular stress corrosion cracking (IGSCC) and fatigue in the boiling water reactor (BWR) environment. Recently, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Therefore, the structural integrity assessment for primary piping taking PWSCC into consideration has become important. This paper details the improvement of PASCAL-SP to evaluate the failure probability of primary pipe taking PWSCC into consideration. We introduce several probabilistic evaluation models such as crack initiation, crack growth and crack detection models related to PWSCC into PASCAL-SP. As numerical examples, the failure probabilities for circumferential and axial cracks in nickel-based alloy weld in pipe in the PWR primary water environment are calculated. We also evaluate the influence of non-destructive inspection on failure probabilities. On the basis of the numerical results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability of primary pipe taking PWSCC into account.

In recent years, increasingly higher strength grades of steel have been used to meet the increasingly demanding requirements placed on rockbolts in underground coal mines. An unfortunate consequence of this has been the occurrence of premature failure of the rockbolts due to stress corrosion cracking. Stress corrosion cracking is a complex phenomenon which involves slow, progressive crack growth under the application of a sustained load (either residual or applied) in what is usually only a mildly corrosive environment. The failure occurs in a brittle manner below the ultimate tensile strength of the material. An understanding of the mechanisms associated with rockbolt failure is imperative in order to develop effective strategies that will reduce the incidence of failure. This paper reports on a forensic investigation directed towards this goal. The work involved fractographic analysis of failed rockbolts retrieved from underground coal mines, identification of the corrosion product in the stress corrosion cracks, and analysis of mine water from affected areas in the mines. It was found that two distinctly different types of mine water were associated with the stress corrosion cracking failures, these being high pH and near neutral pH. The fractographic analysis showed that the micro-mechanism of stress corrosion cracking was quite different for the two types of minewater. Different types of corrosion product were observed within the stress corrosion cracks produced from each of the two different minewaters, providing further confirmation that the stress corrosion cracking mechanism was different in the two cases.

Since the mid-90s, GRS has been dealing systematically with the assessment of the influences of ageing on the safety of nuclear power plants. This also involves the generic evaluation of operating experience. In the following, the procedure, methodological approaches and relevant results are presented in an exemplary manner for passive mechanical components. The results show that the number of events influenced by ageing phenomena that have so far occurred in German plants is low and that measures were taken where necessary to detect, monitor and control any changes due to ageing. As new insights may emerge over time concerning the assessment of the ageing behaviour of the components, it is considered necessary to continue with the ongoing generic evaluation of operating experience in order to obtain timely indications of any safety-relevant changes.

Austenitic stainless steels are usually chosen to make many components of nuclear power plants (NPPs). However, their microstructure in the heat-affected zone (HAZ) will change during the welding process. Some failures of the weld joints, mainly stress corrosion cracking (SCC), have been found to be located in the HAZ. In this research, the microstructure, micro-hardness, residual strain and SCC behavior at different locations of the 316L HAZ cut from a safe-end dissimilar metal weld joint were studied. However, traditional optical microscope observation could not find any microstructural difference between the HAZ and the base metal, higher residual strain and micro-hardness, and higher fraction of random high-angle grain boundaries were found in the HAZ than in the base metal when studied by using electron back-scattering diffraction scanning and micro-hardness test. What’s more, the residual strain, the micro-hardness and the fraction of random grain boundaries decreased, while the fraction of coincidence site lattice grain boundaries increased with increasing the distance from the fusion boundary in 316L HAZ. Creviced bent beam test was applied to evaluate the SCC susceptibility at different locations of 316L HAZ and base metal. It was found that the HAZ had higher SCC susceptibility than the base metal and SCC resistance increased when increasing the distance from the fusion boundary in 316L HAZ.