Study on Application of PFM Analysis Method to Japanese Code for RPV Integrity Assessment Under PTS Events

2015 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Input Data ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Analysis Method ◽  
Integrity Assessment ◽  
Analysis Models ◽  
Reactor Pressure ◽  
The U.S

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

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.

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

2015 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Kazuya Osakabe ◽  
Kentaro Yoshimoto
Keyword(s):  
Fracture Toughness ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Confidence Limits ◽  
Lognormal Distributions ◽  
Integrity Assessment ◽  
Probabilistic Evaluation ◽  
Evaluation Models ◽  
Reactor Pressure

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects

2003 ◽  
Author(s):  
Dominique Moinereau ◽  
Jean-Michel Frund ◽  
Henriette Churier-Bossennec ◽  
Georges Bezdikian ◽  
Alain Martin
Keyword(s):  
Structural Integrity ◽  
Three Dimensional ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Short Description ◽  
Constraint Effect ◽  
Materials Properties ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

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.

Probabilistic Leak Before Break

2018 ◽  
Author(s):  
L. Stefanini ◽  
F. J. Blom
Keyword(s):  
Fracture Toughness ◽  
Yield Stress ◽  
Master Curve ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Leakage Rate ◽  
Critical Crack ◽  
Critical Crack Length ◽  
Integrity Assessment ◽  
Leak Before Break

In this study a probabilistic Leak-Before-Break (LBB) analysis was carried out based on the R6 FAD Option 1 assessment method. The method uses the material fracture toughness and yield stress in order to determine, deterministically, a Critical Crack Length (CCL) and a Leakage Rate (LR) through a crack. In order to define the fracture toughness of the material, the Master Curve approach was used accordingly to BS7910:2013 Annex J. Initially, deterministic analyses were carried out and the fracture toughness and yield stress were set to 190 MPa√m and 158 MPa, respectively. In order to implement a probabilistic approach, the yield stress and fracture toughness were introduced as stochastic parameter. The Fracture toughness was generated using a Weibull distribution to match the Master Curve. The distribution was built such that 190 MPa√m represents the 5% probability fracture toughness. The Yield stress (0.2% proof strength) was generated using a normal distribution with standard deviation 10.35 MPa such that the average value was 175 MPa and the lower bound (5% of probability of occurrence) was 158 MPa. The choice of building the distribution as above mentioned was justified by the fact that in structural integrity assessment the lower 5% is generally used for material parameters. Thus, once a Detectable Leakage Rate (DLR) was determined, it was possible to assign an implicit probability of failure to the deterministic case. The calculations were then extended by using several LR formulas. The calculations were carried out making use of the probabilistic software RAP++ coupled to MATLAB. The probabilities of failure were calculated with regard to a postulated DLR and a DLRSF corrected with a safety factor of 10. The probabilities of failure for the DLRSF were proved to be 9 to 15 times higher than for the postulated DLR case, which leads to the opportunity of conservatism reduction.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During PTS Events

2012 ◽  
Author(s):  
Kunio Onizawa ◽  
Koichi Masaki ◽  
Jinya Katsuyama
Keyword(s):  
Crack Initiation ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Integrity Analysis ◽  
Reactor Pressure

In order to apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), PFM analysis code has been developed at JAEA. Using the PFM analysis code, PASCAL version 3, the conditional probabilities of crack initiation and fracture for an RPV during pressurized thermal shock events have been analyzed. Sensitivity analyses on some input parameters were performed to clarify the effect on the conditional fracture probability. Comparison between the conditional probabilities and temperature margin (ΔTm) from current deterministic analysis method were made for some model plant conditions of domestic typical old-type RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

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