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.

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 reactor pressure vessels during pressurized thermal shock

2008 ◽  
Vol 22 (8) ◽  
pp. 1451-1459 ◽  
Author(s):  
Myung Jo Jhung ◽  
Seok Hun Kim ◽  
Young Hwan Choi ◽  
Sunggyu Jung ◽  
Jong Min Kim ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Good Practice Handbook for PTS Evaluation (Result of the IAEA Co-Ordinated Project on PTS)

2009 ◽  
Author(s):  
Vladislav Pistora ◽  
Milan Brumovsky ◽  
Nigel Taylor
Keyword(s):  
Thermal Shock ◽  
Best Practice ◽  
Structural Integrity ◽  
Good Practice ◽  
Pressure Vessels ◽  
Research Project ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Sensitivity Studies ◽  
Reactor Pressure

Integrity and lifetime of reactor pressure vessels are practically determined by their behavior during “pressurized thermal shock” (PTS) emergency regimes as the most severe regimes during reactor operation. Assessment of these potential regimes is carried out mostly in deterministic way but used procedures are different in different countries. Proper and reliable evaluation of these PTS regimes depends on many parameters and approaches used during computations. During the period 2005 – 2008, the Coordinated Research Project 9 (CRP 9) “Review and Benchmark of Calculation Methods for Structural Integrity Assessment of RPVs During PTS” was organised by the IAEA. The overall objective of this Coordinated Research Project was to perform benchmark deterministic calculations of a typical pressurised thermal shock (PTS) regime and finally to recommend the best practice for PTS assessment. This paper describes main results and collected experience within this project that were bases for the preparation of the “Good Practice Handbook for Deterministic Evaluation of the Integrity of a Reactor Pressure Vessel during a Pressurised Thermal Shock” that will be issued as an IAEA TECDOC. Main parameters discussed in this handbook are: - selection of overcooling sequences; - thermal-hydraulics analyses; - temperature and stress field calculations; - crack tip loading incl.K estimations; - integrity assessment; - analyses of nozzles; - national practices; - results from sensitivity studies. Finally, recommendations for reliable and correct PTS evaluation are given.

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 ....

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.

Development of Probabilistic Fracture Mechanics Analysis Code PASCAL Version 4 for Reactor Pressure Vessels

2018 ◽  
Author(s):  
Kai Lu ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shumpei Uno
Keyword(s):  
Structural Integrity ◽  
Prediction Method ◽  
Pressure Vessels ◽  
Stress Distributions ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Integrity Assessment ◽  
Failure Frequency ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWRs) for structural integrity assessment of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. By reflecting the latest knowledge and findings, the PASCAL code has been continuously improved. In this paper, the development of PASCAL Version 4 (hereafter, PASCAL4) is described. Several analysis functions incorporated into PASCAL4 for evaluating the failure frequency of RPVs are introduced, for example, the evaluation function of confidence level of failure frequency considering epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions and KI calculation methods considering complicated stress distributions, and the recent Japanese irradiation embrittlement prediction method. Finally, using PASCAL4, a PFM analysis example for a Japanese model RPV is presented.

A Comparison Between Probabilistic and Deterministic Fracture Mechanics Assessments of the Structural Integrity of a Reactor Pressure Vessel Subjected to a Pressurized Thermal Shock Transient

2013 ◽  
Author(s):  
Zengliang Gao ◽  
Yuebing Li ◽  
Yuebao Lei
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Safety Factor ◽  
Failure Probability ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Deterministic Approach ◽  
Pressurized Thermal Shock ◽  
Deterministic Methods ◽  
Reactor Pressure

Both probabilistic and deterministic methods are used in structural integrity assessment of reactor pressure vessels (RPV) under pressurized thermal shock (PTS) conditions. The deterministic assessment is normally performed using flaw assessment procedures based on linear elastic or elastic-plastic fracture mechanics. Over the past two decades, the probabilistic assessment approach, which is based on probabilistic fracture mechanics (PFM), has undergone continuous development, mostly driven by the desire to address some of the weaknesses of the deterministic approach and to facilitate increasing the life and safety of nuclear power plants. In this paper, structural integrity assessments for a selected RPV subjected to a typical PTS transient are performed using the deterministic approach according to different flaw assessment codes. The failure probabilities corresponding to the deterministic facture mechanics method with defined safety factors are evaluated and compared with the failure probability value determined using the PFM method. Several sources of uncertainty that affect the assessment of the structural integrity of an RPV under PTS, including uncertainties in the material property values, the fracture toughness and the flaw size are incorporated in the failure probability evaluation. The response distribution of crack driving force is obtained from the PFM analysis and the failure probability is calculated using Monte Carlo simulation, where the failure criteria used in the deterministic assessment are adopted. The results of analysis from the two approaches are compared and discussed. The results show that the defined safety factor in the deterministic methods does affect the limit failure probability implied by the method. However, there is no unique relationship between safety factor and the limit failure probability.

Development of Crack Evaluation Models for Probabilistic Fracture Mechanics Analyses of Japanese Reactor Pressure Vessels

2018 ◽  
Author(s):  
Kai Lu ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Evaluation Models ◽  
Reactor Pressure

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessment of reactor pressure vessels (RPVs). The most recent release is PASCAL Version 4 (hereafter, PSACAL4) which can be used to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and pressurized thermal shock events. For the integrity assessment of RPVs, development of crack evaluation models is important. In this study, finite element analyses are performed firstly to verify the stress intensity factor calculations of cracks in PASCAL4. In addition, the applicability of the crack evaluation models in PASCAL4 such as the location of embedded cracks, crack shape and depth of surface cracks, and the increment of crack propagation is investigated. Based on sensitivity analyses of crack evaluation models for Japanese RPVs using PASCAL4, the effects of these evaluation models on failure frequency are clarified. From the analysis results, crack evaluation models recommended to the failure frequency evaluation for a Japanese model RPV are discussed.

Sign in / Sign up

Export Citation Format

Share Document