An Engineering Methodology for Structural Integrity Assessments Using Probabilistic Fracture Mechanics

1997 ◽  
pp. 1477-1484
Author(s):  
Claudio Ruggieri ◽  
Robert H. Dodds
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

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.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as 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. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese 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 non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, 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.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

2020 ◽  
Vol 143 (2) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Thermal Transients ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

Review of the Current Conservatisms in the Assessment of Thin-Walled Structures

2009 ◽  
Author(s):  
R. S. Kulka ◽  
A. J. Price
Keyword(s):  
Fracture Mechanics ◽  
Best Practice ◽  
Structural Integrity ◽  
Fracture Behaviour ◽  
Stress Distributions ◽  
Limit Loads ◽  
Thin Walled Structures ◽  
Commercial Drivers ◽  
Thin Walled ◽  
Integrity Assessments

Current methods for performing fracture mechanics assessments of thin-walled structures possess significant levels of conservatism, since they are often for general use with both thin-walled and thick-walled structures. In many industries, there are significant commercial drivers for reducing the amount of conservatism in these assessments. An enhanced understanding of the fracture behaviour of thin-walled structures in different situations may lead to the development of more appropriate structural integrity assessments. A review of the information pertaining to fracture mechanics analysis of thin-walled structures has been conducted. There are indications that improvements to the best practice methodology may be made, through improved tensile and toughness properties, consideration of limit loads, and the effect of residual stress distributions.

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.

The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients

2008 ◽  
Author(s):  
Terry Dickson ◽  
Mark EricksonKirk
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Regulatory Commission ◽  
Technical Basis

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.

An Improved Method for Postulating Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluation

2002 ◽  
Author(s):  
F. A. Simonen ◽  
T. L. Dickson
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Reactor Pressure

This paper presents an improved model for postulating fabrication flaws in reactor pressure vessels (RPVs) and for the treatment of measured flaw data by probabilistic fracture mechanics (PFM) codes that are used for structural integrity evaluations. The model used to develop the current pressurized thermal shock (PTS) regulations conservatively postulated that all fabrication flaws were inner-surface breaking flaws. To reduce conservatisms and uncertainties in flaw-related inputs, the United States Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) that has resulted in data on fabrication flaws from non-destructive and destructive examinations of actual RPV material. Statistical distributions have been developed to characterize the number and sizes of flaws in the various material regions of a vessel. The regions include the main seam welds, repair welds, base metal of plates and forgings, and the cladding that is applied to the inner surface of the vessel. Flaws are also characterized as being located within the interior of these regions or along the weld fusion lines that join the regions. Flaws are taken that occur at random locations relative to the embrittled inner region of the vessel. The probabilistic fracture mechanics model associates each of the simulated flaw types with the fracture properties of the region being addressed.

Engineering Criticality Assessments of Floating Offshore Platforms Based on Time Domain Structural Response Analysis

2021 ◽  
Author(s):  
Jeffrey O’Donnell ◽  
Johyun Kyoung ◽  
Sagar Samaria ◽  
Anil Sablok
Keyword(s):  
Fracture Mechanics ◽  
Time Domain ◽  
Structural Integrity ◽  
Structural Response ◽  
Life Extension ◽  
Response Analysis ◽  
Offshore Platforms ◽  
Industry Standards ◽  
Offshore Industry ◽  
Integrity Assessments

Abstract This paper presents a time-domain S-N fatigue analysis and an approach to reliable and robust engineering criticality assessments to supplement or provide an alternative to S-N fatigue assessments of offshore platform structures based on time domain structural response analysis. It also provides recommendations for industry standards to improve guidance for structural integrity assessments of offshore platforms using fracture mechanics. Demand continues to grow in the offshore industry to attain value from captured operational data for a number of purposes, including the reduction of uncertainties in structural integrity assessments during design and over the operational lifetime of floating offshore platforms. Recent advances in time domain structural analysis technology demonstrate substantially more accurate assessments of non-linear platform loadings and responses with enhanced computational efficiency. The current S-N approach for fatigue design and integrity assessments calculates a fatigue damage factor that does not address how loading occurs over time (ABS, DNVGL-RP-C203). For the present study, engineering criticality assessments (ECAs) based on fracture mechanics theory (BS 7910) are applied utilizing time-domain loading information theory. The ECA returns the smallest initial flaws that can grow to a critical size during a design lifetime, which can serve as an indicator of acceptability during design, a technical basis for in-service inspection intervals and facilitates asset integrity and life extension assessments. Critical initial flaws are calculated using the Paris Law (BS 7910) and cumulative fatigue crack growth in two ways: with and without an integrated and consistent check for fracture instability. The results are compared with those from S-N fatigue analyses and recommendations are provided.

Benchmarking Study on Probabilistic Fracture Mechanics Analysis Codes xLPR and PASCAL-SP Considering Primary Water Stress Corrosion Cracking

2021 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Water Stress ◽  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Probabilistic Fracture Mechanics ◽  
Primary Water ◽  
Failure Probabilities

Abstract Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.

Probabilistic Fracture Mechanics Application for Alloy 82/182 Welds in PWRS

2010 ◽  
Author(s):  
Jong-Dae Hong ◽  
Changheui Jang
Keyword(s):  
Fracture Mechanics ◽  
Water Chemistry ◽  
Power Plants ◽  
Structural Integrity ◽  
Boundary Carbide ◽  
Pressurized Water ◽  
Butt Weld ◽  
Monte Carlo Simulation Technique ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

In operating PWRs (Pressurized Water Reactors), incidents of Alloy 82/182 cracking increased the concern for structural integrity of butt weld locations recently, because of high weld residual stresses. Studies on PWSCC (Primary Water Stress Corrosion Cracking) have been mainly performed using deterministic approaches by controlling parameters, but a quantitative evaluation is difficult because of large uncertainties in each parameter and test results. The purposes of this paper are to provide a probabilistic fracture mechanics (PFM) analysis methodology and quantify failure probabilities for Alloy 82/182 welds in primary piping systems of nuclear power plants. To calculate failure probabilities, Monte Carlo simulation technique was used. To estimate the time to crack initiation, material susceptibility was quantified considering the effects of various processing, grain boundary carbide coverage, water chemistry including zinc addition, and so on. In crack growth analysis, crack orientation and the effects of water chemistry including dissolved hydrogen concentration were considered. And the effects of weld repair were evaluated.

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