Fitness-for-service NDE and fracture mechanics for pressure vessels and piping

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.

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Optimizing Probe Active Aperture for Phased Array Weld Inspections

Materials Evaluation ◽

10.32548/2021.me-04220 ◽

2021 ◽

Vol 79 (8) ◽

pp. 797-804

Author(s):

Anmol Birring

Keyword(s):

Fracture Mechanics ◽

Ultrasonic Testing ◽

Phased Array ◽

Pressure Vessels ◽

Acceptance Criteria ◽

Nondestructive Technique ◽

Cross Sectional ◽

Area Of Interest ◽

Array Technology ◽

Selection Of

Phased array ultrasonic testing (PAUT) has become a popular nondestructive technique for weld inspections in piping, pressure vessels, and other components such as turbines. This technique can be used both in manual and automated modes. PAUT is more attractive than conventional angle-beam ultrasonic testing (UT), as it sweeps the beam through a range of angles and presents a cross-sectional image of the area of interest. Other displays are also available depending on the software. Unlike traditional A-scan instruments, which require the reconstruction of B- and C-scan images from raster scanning, a phased array image is much simpler to produce from line scans and easier to interpret. Engineering codes have incorporated phased array technology and provide steps for standardization, scanning, and alternate acceptance criteria based on fracture mechanics. The basis of fracture mechanics is accurate defect sizing. There is, however, no guidance in codes and standards on the selection and setup of phased array probes for accurate sizing. Just like conventional probes, phased array probes have a beam spread that depends on the probe’s active aperture and frequency. Smaller phased array probes, when used for thicker sections, result in poor focusing, large beam spread, and poor discontinuity definition. This means low resolution and oversizing. Accurate sizing for fracture mechanics acceptance criteria requires probes with high resolution. In this paper, guidance is provided for the selection of phased array probes and setup parameters to improve resolution, definition, and sizing of discontinuities.

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Practical Fracture Mechanics Applications to Design of High Pressure Vessels

10.21236/ada036667 ◽

1976 ◽

Cited By ~ 3

Author(s):

Thomas E. Davidson ◽

Joseph F. Throop

Keyword(s):

High Pressure ◽

Fracture Mechanics ◽

Pressure Vessels

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

Volume 1B: Codes and Standards ◽

10.1115/pvp2017-66004 ◽

2017 ◽

Cited By ~ 1

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.

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NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

ASME 2011 Pressure Vessels and Piping Conference: Volume 6, Parts A and B ◽

10.1115/pvp2011-57112 ◽

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.

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

Volume 3: Design and Analysis ◽

10.1115/pvp2008-61392 ◽

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.

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Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

Volume 1: Codes and Standards ◽

10.1115/pvp2014-28621 ◽

2014 ◽

Cited By ~ 1

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.

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Uncertainties in Pressurized Thermal Shock Analyses

Volume 3: Design and Analysis ◽

10.1115/pvp2019-94076 ◽

2019 ◽

Cited By ~ 1

Author(s):

M. Niffenegger ◽

O. Costa Garrido ◽

D. F. Mora ◽

G. Qian ◽

R. Mukin ◽

...

Keyword(s):

Fracture Mechanics ◽

Thermal Shock ◽

Probabilistic Approach ◽

Pressure Vessels ◽

Pressurized Water Reactor ◽

Ansys Fluent ◽

Pressurized Water ◽

Integrity Assessment ◽

Safety Margins ◽

Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

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Verification of Probabilistic Fracture Mechanics Analysis Code Through Benchmark Analyses

Volume 1B: Codes and Standards ◽

10.1115/pvp2018-84963 ◽

2018 ◽

Author(s):

Yinsheng Li ◽

Shumpei Uno ◽

Koichi Masaki ◽

Jinya Katsuyama ◽

Terry Dickson ◽

...

Keyword(s):

Fracture Mechanics ◽

The United States ◽

Pressure Vessels ◽

Irradiation Embrittlement ◽

Energy Agency ◽

Probabilistic Fracture Mechanics ◽

Failure Frequency ◽

Pressurized Thermal Shock ◽

Great Confidence ◽

Reactor Pressure

A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency based on Japanese methods and data to evaluate failure probabilities and failure frequencies of Japanese reactor pressure vessels (RPVs) considering pressurized thermal shock (PTS) events and neutron irradiation embrittlement. To verify PASCAL, we have been performing benchmark analyses by using a PFM code FAVOR which has been developed in the United States and utilized in nuclear regulation. Based on two-year activities, the applicability of PASCAL in failure probability and failure frequency evaluation of Japanese RPVs was confirmed with great confidence. The analysis conditions, approaches and results are given in this paper.

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