Assessment of Large-Scale Pressurized Thermal Shock Experiments Using the FAVOR Fracture Mechanics Computer Code

2004 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. Kirk
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Large Scale ◽  
Major Element ◽  
Computer Code ◽  
Fracture Analysis ◽  
Pressure Vessels ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Evaluation Project

Large-scale experiments of pressure vessels performed at the Oak National Laboratory (ORNL)1 in the mid 1980s validated the applicability of the linear-elastic fracture mechanics (LEFM) computational methodology for application to fracture analysis of reactor pressure vessels (RPVs) in nuclear power plants. The current federal regulations to insure that nuclear RPVs maintain their structural integrity, when subjected to transients such as pressurized thermal shock (PTS) events, were derived in the early-mid 1980s from a comprehensive computational methodology of which LEFM is a major element. Recently, the United States Nuclear Regulatory Commission (USNRC) has conducted the PTS re-evaluation project that has the objective to establish a technical basis for a potential relaxation to the current PTS regulations which could have profound implications for plant license-extension considerations. The PTS re-evaluation project has primarily consisted of the development and application of an updated risk-based computational methodology that has been implemented into the Fracture Analysis of Vessels: Oak Ridge (FAVOR) computer code. LEFM continues to be a major element of the updated computational methodology. As part of the PTS re-evaluation program, there has been an extensive effort to validate that FAVOR has an accurate implementation of the LEFM methodology. This effort has consisted of the successful benchmarking of thermal analysis, stress analysis, and LEFM fracture analysis results between FAVOR and ABAQUS, a commercial general-purpose finite element computer code that has fracture mechanics capabilities, for a range of transient descriptions. The NRC has also participated in international round-robin benchmarking exercises in which FAVOR-generated solutions to well-specified PTS problems have been compared to solutions generated by other research institutions. A more fundamental aspect of the ongoing validation of FAVOR is demonstration that FAVOR can be used to successfully predict the results of large-scale fracture experiments. The objective of this paper is to document the FAVOR analysis of the first large-scale pressurized thermal shock experiment (PTSE) performed at ORNL. Results of these analyses provide validation that FAVOR accurately predicts the cleavage fracture initiation of a long surface breaking flaw in a large-scale thick-walled pressure vessel.

The Impact of an Improved Flaw Model on a Pressurized Thermal Shock Evaluation

2002 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Computational Models ◽  
The United States ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Evaluation Program ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
The Impact ◽  
Flaw Characterization

The current regulations for pressurized thermal shock (PTS) were derived from computational models that were developed in the early-mid 1980s. The computational models utilized in the 1980s conservatively postulated that all fabrication flaws in reactor pressure vessels (RPVs) were inner-surface breaking flaws. It was recognized at that time that flaw-related data had the greatest level of uncertainty of the inputs required for the probabilistic-based PTS evaluations. To reduce this uncertainty, the United States Nuclear Regulatory Commission (USNRC) has in the past few years supported research at Pacific Northwest National Laboratory (PNNL) to perform extensive nondestructive and destructive examination of actual RPV materials. Such measurements have been used to characterize the number, size, and location of flaws in various types of welds and the base metal used to fabricate RPVs. The USNRC initiated a comprehensive project in 1999 to re-evaluate the current PTS regulations. The objective of the PTS Re-evaluation program has been to incorporate advancements and refinements in relevant technologies (associated with the physics of PTS events) that have been developed since the current regulations were derived. There have been significant improvements in the computational models for thermal hydraulics, probabilistic risk assessment (PRA), human reliability analysis (HRA), materials embrittlement effects on fracture toughness, and fracture mechanics methodology. However, the single largest advancement has been the development of a technical basis for the characterization of fabrication-induced flaws. The USNRC PTS-Revaluation program is ongoing and is expected to be completed in 2002. As part of the PTS Re-evaluation program, the updated risk-informed computational methodology as implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, including the improved PNNL flaw characterization, was recently applied to a domestic commercial pressurized water reactor (PWR). The objective of this paper is to apply the same updated computational methodology to the same PWR, except utilizing the 1980s flaw model, to isolate the impact of the improved PNNL flaw characterization on the PTS analysis results. For this particular PWR, the improved PNNL flaw characterization significantly reduced the frequency of RPV failure, i.e., by between one and two orders of magnitude.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
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.

scholarly journals Probabilistic Fracture Mechanics Round Robin Analysis of Reactor Pressure Vessels during Pressurized Thermal Shock

2010 ◽  
Vol 47 (12) ◽  
pp. 1131-1139 ◽  
Author(s):  
Myung Jo JHUNG ◽  
Seok Hun KIM ◽  
Young Hwan CHOI ◽  
Yoon Suk CHANG ◽  
Xiangyuan XU ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Round Robin ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

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.

An Overview of the Pressurized Thermal Shock Re-Evaluation Project

2005 ◽  
Author(s):  
Terry L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Thermal Shock ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Technical Basis ◽  
Evaluation Project

In 1999, a study sponsored by the United States Nuclear Regulatory Commission (NRC) suggested that advances in the technologies associated with the physics of pressurized-thermal-shock (PTS) events developed since the derivation of the PTS regulations (established in the early-mid eighties) had the potential to establish a technical basis that could justify a relaxation in the current PTS-related regulations. A relaxation of these regulations could have profound implications for plant license extension considerations. Subsequently, the NRC initiated the interdisciplinary PTS Re-evaluation Project. During the five year project, an updated comprehensive computational methodology evolved, within the framework established by modern probabilistic risk assessment (PRA) techniques, through interactions among experts in relevant disciplines from the NRC staff, their contractors, and representatives from the nuclear industry. During 2004, the updated computational methodology was applied to three domestic commercial pressurized water reactors (PWRs). The most recent results of the PTS Re-evaluation Project provide a technical basis to support a relaxation of the current PTS regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. The details of the updated computational methodology, the mathematical models, the analysis results, key findings, and supporting information have recently been drafted in several very detailed and lengthy formal reports. These reports are currently under review at the NRC. An objective of this paper is to provide a short overview of the improved computational methodology, analysis results, and key findings of the PTS re-evaluation project. To demonstrate that a technical basis has been established to support a relaxation of the current PTS regulations, it is helpful to understand the derivation of the current PTS regulations; therefore, another objective of this paper is to contrast the interpretation of the analysis results of the PTS re-evaluation to those performed in the eighties from which the current PTS regulations were derived.

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.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

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