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.

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.

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

2015 ◽  
Author(s):  
Gary L. Stevens ◽  
Mark T. Kirk ◽  
Terry Dickson
Keyword(s):  
Fracture Mechanics ◽  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Stress Concentrations ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Regulatory Issue ◽  
Reactor Pressure

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

Distributions of Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluations

2002 ◽  
Author(s):  
F. A. Simonen ◽  
G. J. Schuster ◽  
S. R. Doctor ◽  
T. L. Dickson
Keyword(s):  
Pacific Northwest ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Distribution Functions ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Reactor Pressure

To reduce uncertainties in flaw-related inputs for probabilistic fracture mechanics (PFM) evaluations, the U.S. Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) involving nondestructive and destructive examinations for fabrication flaws in reactor pressure vessel (RPV) material. Using these data, statistical distributions have been developed to characterize the flaws in regions of a RPV. The regions include the main seam welds, repair welds, base metal, and the cladding at the inner surface of the vessel. This paper summarizes the available data and describes the treatment of these data to estimate flaw densities, flaw depth distributions, and flaw aspect ratio distributions. The methodology has generated flaw-related inputs for PFM calculations that have been part of an effort to update pressurized thermal shock (PTS) regulations. Statistical treatments of uncertainties in the parameters of the flaw distribution functions are part of the inputs to the PFM calculations. The paper concludes with a presentation of some example input files that have supported evaluations by USNRC of the risk of vessel failures caused by PTS events.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

2017 ◽  
Author(s):  
Jinya Katsuyama ◽  
Kazuya Osakabe ◽  
Shumpei Uno ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Neutron Irradiation ◽  
Structural Integrity ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the standards developed by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the statistical distributions of various influence factors related to ageing due to the long-term operation. At Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, JAEA has developed a guideline for the PFM based structural integrity assessment of RPVs to promote the applicability of PFM in Japan and achieve the objective that an engineer/analyst who familiar with the fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body (general requirements), explanation (guidance), and several supplements. The technical basis for PFM analysis is also provided, and the new information and better fracture mechanics models are included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

2020 ◽  
Vol 142 (2) ◽  
Author(s):  
Jinya Katsuyama ◽  
Kazuya Osakabe ◽  
Shumpei Uno ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Neutron Irradiation ◽  
Structural Integrity ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

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 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.

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.

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

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.

Verification of Probabilistic Fracture Mechanics Analysis Code Through Benchmark Analyses

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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