scholarly journals J-integral elastic plastic fracture mechanics evaluation of the stability of cracks in nuclear reactor pressure vessels

1980 ◽  
Author(s):  
M. P. Gomez ◽  
R. M. McMeeking ◽  
D. M. Parks
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
J Integral ◽  
Elastic Plastic ◽  
The Stability ◽  
Reactor Pressure

FAVOR Version 16.1: A Computer Code for Fracture Mechanics Analyses of Nuclear Reactor Pressure Vessels

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Computer Code ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Current Status ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life. Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a). This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

Computational Aspect of Nonlinear Fracture Mechanics Application

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Andrii Oryniak
Keyword(s):  
Residual Stress ◽  
Fracture Mechanics ◽  
Thermal Stresses ◽  
Crack Tip ◽  
Pressure Vessels ◽  
Point Of View ◽  
Computational Aspect ◽  
J Integral ◽  
Elastic Plastic ◽  
Plastic Analysis

The development of powerful commercial computer programs made the concept of J-integral as computational parameter of fracture mechanics to be a very attractive one. It is equivalent to SIF in linear case, it converges in numerical calculation and the same results are obtained by different codes (programs). Besides, it is widely thought that elastic-plastic analysis gives bigger values than elastic SIF ones what is good from regulatory point of view. Such stand was reflected in the recommended by IAEA TECDOC 1627 (February 2010) devoted to pressurized thermal shock analysis of reactor pressure vessels, where the embedded crack in FEM mesh, elastic-plastic analysis with simultaneous determination of J-integral was stated as the best option of analysis. But at that time all the most widely used software were not able to treat the residual stresses, the thermal stresses in case of two different materials. Such a contradiction between requirements and the possibilities made a lot of problems for honest contractors especially in countries where the regulator had no own experience in calculation and completely relied on the authority of international documents. This means that at that time the said recommendations were harmful. The main reason of such a situation was the absence of the carefully elaborated examples. Now the capabilities and accuracy of such software are increasing. Nevertheless, some principal ambiguities and divergences of computations results in various J-integral contours around the crack tip still exist. They are exhibited when the large plastic zone emerges at the crack tip. Other problem is influence of the history of loading and the specification of the time of crack insertion in the mesh including the time of emergence of residual stress. This paper is invitation for discussion of the accuracy and restriction of computational J-integral. With this aim the detailed analysis of some simplified 2D examples of calculation of elastic -plastic J-integral for surface crack with accounting for residual stress, thermal stress and inner pressure are performed and commented. The attainment of consensus among the engineering society for treating the outcome results is the prerequisite for practical application of computational elastic plastic J-integral.

Coupling of Neutron Transport and Probabilistic Fracture Mechanics Codes for Analysis of Embrittled Reactor Pressure Vessels

2020 ◽  
Author(s):  
Benjamin W. Spencer ◽  
William M. Hoffman ◽  
Benjamin S. Collins ◽  
Shane C. Henderson
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Radiation Transport ◽  
Neutron Transport ◽  
Critical Role ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Significant Advance ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

Abstract In light water nuclear reactors, the reactor pressure vessel (RPV) plays the critical role of containing the reactor and coolant and is expected to maintain its integrity under a variety of conditions. The RPV is subjected to high neutron flux and temperature, which lead to material embrittlement and increase its susceptibility to fracture. The degree of embrittlement is highly dependent on the local fluence, so it is critical to accurately characterize the spatial distribution of fluence in the RPV when performing a probabilistic fracture mechanics (PFM) analysis, which individually evaluates the flaws introduced in the manufacturing process. Radiation transport codes typically focus on computing the neutron distribution within the nuclear reactor core to understand phenomena relevant to reactor physics. However, recent developments in the Virtual Environment for Reactor Applications (VERA) allow it to solve for a fuel-pin resolved in-core fission source and use Monte Carlo methods to compute neutron fluence accumulation at locations away from the core, such as within the RPV, with unprecedented accuracy. Grizzly, a PFM code, can read in a 3D map of the fluence computed by VERA, and directly use that in the PFM analysis. This represents a significant advance in the accuracy of the fluence used in PFM calculations. This paper provides an overview of this modeling approach demonstrates its application on PFM analyses of representative RPVs.

Evaluation of Cracks Embedded in Zones With Large Scale Plasticity in Subsea HPHT Equipment – Comparing FAD and Driving Force Approaches

2021 ◽  
Author(s):  
Mandar Kulkarni ◽  
Carlos Lopez ◽  
Daniel Kluk ◽  
John Chappell
Keyword(s):  
Fracture Mechanics ◽  
Plastic Strain ◽  
Driving Force ◽  
Large Scale ◽  
Plastic Region ◽  
Assessment Method ◽  
Pressure Vessels ◽  
Integral Approach ◽  
J Integral ◽  
Elastic Plastic

Abstract Fracture mechanics assessments for pressure vessels are performed to determine critical flaw sizes and/or estimate the fatigue life of a growing crack as a means of establishing inspection intervals for the equipment. In most cases the evaluation is performed based on methods described in API 579-1/ASME FFS-1 and BS7910. The approaches described in these standards are mostly based on a linear elastic fracture mechanics approach. Even though plasticity can be accounted for by using a failure assessment diagram (FAD); however, even with this approach the effect of plastic strain around the crack is not explicitly considered. This paper presents an approach as per API 579, Annex 9G.5 which recommends utilizing a driving force method whereby the J-integral is directly evaluated from an elastic-plastic finite element model. The main goal is to study differences between the FAD approach against the elastic-plastic J-integral approach wherein the crack is modeled explicitly. Simplified representative geometries are considered for this study. Two scenarios for the plastic zone are considered a) crack present during initial loading with no residual plastic strain and b) crack in a residual stress zone. Different crack sizes are considered for this comparison study ranging from small cracks completely embedded within the plastic region and larger cracks with partial embedment. The paper presents comparison studies which highlight the key differences between different analysis approaches with the aim of identifying the most conservative assessment method for different crack geometries.

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.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

Prediction of Brittle Fracture Under Generalised Elastic Plastic Loading

2008 ◽  
Author(s):  
S. J. Lewis ◽  
C. E. Truman ◽  
D. J. Smith
Keyword(s):  
Fracture Mechanics ◽  
Brittle Fracture ◽  
Ferritic Steel ◽  
Failure Prediction ◽  
J Integral ◽  
Local Approach ◽  
Elastic Plastic ◽  
Failure Data ◽  
Stress States ◽  
Plastic Loading

This article describes an investigation into the ability of a number of different fracture mechanics approaches to predict failure by brittle fracture under general elastic/plastic loading. Data obtained from C(T) specimens of A508 ferritic steel subjected to warm pre-stressing and side punching were chosen as such prior loadings produce considerably non-proportionality in the resulting stress states. In addition, failure data from a number of round notched bar specimens of A508 steel were considered for failure with and without prior loading. Failure prediction, based on calibration to specimens in the as received state, was undertaken using two methods based on the J integral and two based on local approach methodologies.

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