Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP Regarding Stress Corrosion Cracking in Nickel Based Alloy Weld Joint of Piping System in Boiling Water Reactor

2021 ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion Cracking ◽  
Failure Probability ◽  
Weld Joint ◽  
Structural Integrity ◽  
Boiling Water ◽  
Piping System ◽  
Integrity Assessment ◽  
Weld Joints ◽  
Nickel Based Alloy

Abstract In the past few decades, the cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints welded using nickel based alloy in piping system of boiling water reactors. Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a probabilistic structural integrity assessment of weld joint in pipe in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint welded using nickel based alloy in boiling water reactor susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regard to Primary Water Stress Corrosion Cracking

2017 ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Failure Probability ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Integrity Assessment ◽  
Primary Water ◽  
Nickel Based Alloy ◽  
Failure Probabilities

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.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regardto PWSCC

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Crack Detection ◽  
Structural Integrity ◽  
Energy Agency ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, 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). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

2011 ◽  
Author(s):  
Shengjun Yin ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Fracture Mechanics ◽  
Large Scale ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pressure Vessels ◽  
Cooperative Action ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Feasibility Test ◽  
Reactor Pressure

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

2014 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Analysis Method ◽  
Integrity Assessment ◽  
The U.S

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration. In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

The Experimental Work in Support to ATLAS+ Project

2018 ◽  
Author(s):  
Dominique Moinereau ◽  
Patrick Le Delliou ◽  
Anna Dahl ◽  
Yann Kayser ◽  
Szabolcs Szavai ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Experimental Work ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Ductile Material ◽  
Small Scale ◽  
Term Operation ◽  
Integrity Assessment

The 4-years European project ATLAS+ project was launched in June 2017. Its main objective is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. The transferability of ductile material properties from small scale fracture mechanics specimens to large scale components is one of the topics of the project. A large programme of experimental work is to be conducted in support of the development and validation of advanced tools for structural integrity assessment within the framework of the work-package 1 (WP 1): Design and execution of simulation oriented experiments to validate models at different scales. The experimental work is based on a full set of fracture mechanics experiments conducted on standard specimens and large scale components (several pipes and one mock-up), including a full materials characterization. Three materials are considered: • a ferritic steel 15NiCuMoNb5 (WB 36) • an aged austenitic stainless steel weld • a VVER (eastern PWR) dissimilar metal weld (DMW) The paper presents the WP 1, the experimental programme and summarizes the first results.

Study on Weld Residual Stress and Crack Propagation Evaluations for a Saddle-Shaped Weld Joint

2013 ◽  
Author(s):  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Kunio Onizawa
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Weld Joint ◽  
Power Plants ◽  
Structural Integrity ◽  
Piping System ◽  
Stress Distributions ◽  
Analysis Model ◽  
Element Analysis ◽  
Weld Residual Stress

Stress corrosion cracking (SCC) have been observed in reactor coolant pressure boundary piping system at nuclear power plants. When an SCC is found, the structural integrity of piping should be assessed according to a fitness-for-service rule. However, the rule stipulates the assessment procedures for crack growth and failure only for a simple structure such as cylindrical or plate-wise structure. At the present, the methodology even of an SCC growth evaluation for a geometrically complicated piping such as saddle-shaped weld joints has not been established yet. This may be because analyses on the weld residual stress distribution which affects the SCC growth behavior around such portion are difficult to conduct. In this study, we established a finite element analysis model for a saddle-shaped weld joint of pipes. The residual stress distributions produced by the tungsten inert gas (TIG) welding were calculated based on thermal-elastic-plastic analysis with moving and simultaneous heat source models. Analysis results showed complicated weld residual stress distributions, i.e., residual stresses in both hoop and radial directions were tensile at the inner surface near the nozzle corner in branching pipe. SCC growth simulation based on S-version finite element method (S-FEM) using the weld residual stress distributions in saddle-shaped weld joint was also performed. We confirmed an applicability and the accuracy of S-FEM to saddle-shaped weld joint.

Probabilistic Fracture Mechanics Assessment of Wall-Thinned Nuclear Piping System

2005 ◽  
Author(s):  
Sang-Min Lee ◽  
Yoon-Suk Chang ◽  
Young-Jin Kim
Keyword(s):  
Fracture Mechanics ◽  
Probabilistic Approach ◽  
Bending Moment ◽  
Simulation Method ◽  
Combined Loading ◽  
Probabilistic Assessment ◽  
Piping System ◽  
Integrity Assessment ◽  
Assessment Program ◽  
Reliability Method

The integrity of nuclear piping systems must be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc, are required. Up to now, the integrity assessment has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approach is considered as an appropriate method for piping system evaluation. The objectives of this paper are to develop a probabilistic assessment program using reliability index and simulation technique and to estimate the failure probability of wall-thinned pipes in secondary systems. The probabilistic assessment program consists of three evaluation modules that are: first order reliability method, second order reliability method and crude Monte Carlo simulation method. The developed program has been applied to evaluate failure probabilities of wall-thinned piping system subjected to internal pressure, global bending moment and combined loading. The evaluation results showed a promising applicability of the probabilistic integrity assessment program.

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