Failure Assessment of Nuclear Piping Components Due to Combined Degradation Mechanisms by an Advanced Probabilistic Fracture Mechanics Code

Author(s):  
Debashis Datta ◽  
Changheui Jang

Probabilistic failure analysis of nuclear piping components due to a combined degradation mechanisms is a challenging issue. At present the majority of analyses were done by assuming a single failure mechanism for a specific location of a piping system. But in reality, this might not be an accurate approach. A tiny crack might be present in a weld location due to fabrication defect or initiated due to fatigue after a short incubation time of plant’s start up. This pre-existing or initiated crack then may be further matured by the synergistic effect of different probable degradation mechanisms e.g. fatigue, stress corrosion cracking, etc. In this study the development process of an advanced probabilistic fracture mechanics code has been described which can handle this combined failure mechanisms. Numerical examples are also presented to rationalize the use of such methodology.

Author(s):  
David W. Beardsmore ◽  
Karen Stone ◽  
Huaguo Teng

Deterministic Fracture Mechanics (DFM) assessments of structural components (e.g. pressure vessels and piping used in the nuclear industry) containing defects can usually be carried out using the R6 procedure. The aim of such an assessment is to demonstrate that there are sufficient safety margins on the applied loads, defect size and fracture toughness for the safe continual operation of the component. To ensure a conservative assessment is made, a lower-bound fracture toughness, and upper-bound defect sizes and applied loads are used. In some cases, this approach will be too conservative and will provide insufficient safety margins. Probabilistic Fracture Mechanics (PFM) allow a way forward in such cases by allowing for the inherent scatter in material properties, defect size and applied loads explicitly. Basic Monte Carlo Methods (MCM) allow an estimate of the probability of failure to be calculated by carrying out a large number of fracture mechanics assessments, each using a random sample of the different random variables (loads, defect size, fracture toughness etc). The probability of failure is obtained by counting the proportion of simulations which lead to assessment points that lie outside the R6 failure assessment curve. This approach can give good results for probabilities greater than 10−5. However, for smaller probabilities, the calculation may be inefficient and a very large number of assessments may be necessary to obtain an accurate result, which may be prohibitive. Engineering Reliability Methods (ERM), such as the First Order Reliability method (FORM) and the Second Order Reliability Method (SORM), can be used to estimate the probability of failure in such cases, but these methods can be difficult to implement, do not always give the correct result, and are not always robust enough for general use. Advanced Monte Carlo Methods (AMCM) combine the two approaches to provide an accurate and efficient calculation of probability of failure in all cases. These methods aim to carry out Importance Sampling so that only assessment points that lie close to or outside the failure assessment curve are calculated. Two methods are described in this paper: (1) orthogonal sampling, and (2) spherical sampling. The power behind these methods is demonstrated by carrying out calculations of probability of failure for semi-elliptical, surface breaking, circumferential cracks in the inside of a pressure vessel. The results are compared with the results of Basic Monte Carlo and Engineering Reliability calculations. The calculations use the R6 assessment procedure.


2006 ◽  
Vol 306-308 ◽  
pp. 691-696 ◽  
Author(s):  
Indera Sadikin ◽  
Djoko Suharto

The dynamic nature of marine environment is the major cause of fatigue failures in subsea gas pipeline structures. Since the severest loaded part of piping system is the free span, freespan inspection is performed periodically to ensure that no free span exceeds its critical length. The objective of this paper is to optimize free-span inspection interval by means of probabilistic fracture mechanics analysis. Simulation data is taken from previous work of Tronskar [1]. Stress intensity factors at the crack tip are calculated by crack closure technique. Fatigue crack growth is simulated by cycle-by-cycle integration technique. The fracture mechanics analysis is then expanded to probabilistic analysis to include stochastic input parameters. Probability of failure is computed by modified direct simulation method. Based on the result of direct simulation, the studied pipelines are recommended to be inspected every 3 years to make sure that no free span exceeds 30 m.


2021 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Yinsheng Li

Abstract The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.


Author(s):  
Changheui Jang ◽  
Debashis Datta ◽  
Jun-Seog Yang

Recently, the use of risk information in regulatory decision as well as operation and maintenance of nuclear power plant has been gradually increased. Application of risk information to optimize the inservice inspection of nuclear piping system has been one of such an attempt in Korea. An accurate evaluation of the risk, or failure probability for each pipe subcomponent is essential to the application of the risk-informed inservice inspection. For this purpose, a probabilistic fracture mechanics code based on Monte Carlo simulation techniques was developed to handle various flaw shapes and orientations in nuclear piping system. Then, the failure probabilities of various subcomponents of the main coolant loop in a pressurized water reactor were calculated, and the relative risk ranking was determined using the plant’s data. Also, the effects of in-service inspection and frequency on the risk were investigated. Finally, the application of the probabilistic fracture mechanics methods to the risk-informed inservice inspection of nuclear piping system was discussed.


Author(s):  
Yu-Yu Shen ◽  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Ru-Feng Liu

In recent years, the probabilistic fracture mechanics (PFM) approach has been widely applied to estimate the fracture risk of nuclear power plant piping systems. In the paper, the probabilistic fracture mechanics code, PRO-LOCA, developed by the Probabilistic Analysis as a Regulatory Tool for Risk Informed Decision Guidance (PARTRIDGE) project, is employed to practically evaluate the fracture probability of the recirculation piping system welds in a Taiwan domestic boiling water reactor (BWR) nuclear power plant. To begin with, the models based on the real situation of the recirculation piping welds are built. Then, the probabilities of through-wall cracking, leak with different rates, and rupture on the welds considering both in-service inspection and leak detection are analyzed. Meanwhile, the effects of probability of detection curves of ISI on the piping are simulated. Further, the efficiencies of performing the induction heating stress improvement and weld overlay are also studied and discussed. The present work could provide a reference of operation, inspection and maintenance for BWR plants in Taiwan.


1993 ◽  
Vol 8 (1) ◽  
pp. 43-56 ◽  
Author(s):  
Hiroshi Ishikawa ◽  
Akira Tsurui ◽  
Hiroaki Tanaka ◽  
Hidetoshi Ishikawa

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