Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability

2004 ◽  
Author(s):  
Katsuyuki Shibata ◽  
Kunio Onizawa ◽  
YinSheng Li ◽  
Yasuhiro Kanto ◽  
Shinobu Yoshimura
Keyword(s):  
Case Studies ◽  
Failure Probability ◽  
Pressure Vessels ◽  
Aspect Ratios ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Order Of Magnitude ◽  
The Difference ◽  
Reactor Pressure

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.

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.

Improvements to PFM Analysis Code PASCAL and Some Case Studies on RPV Integrity During Pressurized Thermal Shock

2006 ◽  
Author(s):  
Kunio Onizawa ◽  
Katsuyuki Shibata ◽  
Kazuya Osakabe ◽  
Kazuhisa Tanaka
Keyword(s):  
Thermal Shock ◽  
Case Studies ◽  
Crack Detection ◽  
The Monte Carlo Method ◽  
Pressurized Thermal Shock ◽  
Order Of Magnitude ◽  
Vessel Material ◽  
Existence Probability ◽  
Reactor Pressure ◽  
Failure Probabilities

The probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed at JAEA to evaluate the conditional failure probability of a reactor pressure vessel (RPV) containing a flaw under transient conditions such as pressurized thermal shock (PTS). Some functions of PASCAL have been improved such as the Monte Carlo method, the probability of crack detection (POD) by inspection and the graphical user interface (GUI). Using the improved PASCAL, case studies on some parameters for some typical PTS transients have been performed. When the crack existence probability for the Marshall- and PNNL-type distributions is considered, the failure probabilities are approximately the same order of magnitude. The effect of chemical composition of the vessel material was studied through a comparison of conditional failure probabilities for some transients. The effect of POD for inspection was also compared and evaluated in terms of the inspection effectiveness and frequency.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

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.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL Through Benchmark Analyses With FAVOR

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Shumpei Uno ◽  
Jinya Katsuyama ◽  
Terry Dickson ◽  
Mark Kirk
Keyword(s):  
Fracture Mechanics ◽  
The United States ◽  
Pressure Vessels ◽  
Japan Atomic Energy Agency ◽  
Energy Agency ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
Great Confidence ◽  
Reactor Pressure

A probabilistic fracture mechanics (PFM) analysis code called PASCAL has been developed by the Japan Atomic Energy Agency to evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on Japanese data and Japanese methods published for or provided in Japanese codes and standards. To verify this code, benchmark analyses were carried out using the FAVOR code, which was developed in the United States and has been utilized in nuclear regulation. The results of these analyses confirmed with great confidence the applicability of PASCAL to failure probability and frequency evaluation of Japanese RPVs. An outline of PASCAL, the benchmark analysis conditions and analysis results are reported in this paper.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

2009 ◽  
Author(s):  
Vikram Marthandam ◽  
Timothy J. Griesbach ◽  
Jack Spanner
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Historical Perspective ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Potential Benefits ◽  
Vessel Integrity ◽  
Technical Basis ◽  
Reactor Pressure

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

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