Statistical Analyses for Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Building a Master Curve on an Extract of the “Euro” Fracture Toughness Dataset, Controlling Statistical Uncertainty for Both Mono-Temperature and Multi-Temperature Tests

2006 ◽  
Author(s):  
Florent Josse ◽  
Yannick Lefebvre ◽  
Patrick Todeschini ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Ferritic Steels ◽  
Statistical Uncertainty ◽  
Master Curve Method ◽  
Curve Method ◽  
Reactor Pressure

Assessing the structural integrity of a nuclear Reactor Pressure Vessel (RPV) subjected to pressurized-thermal-shock (PTS) transients is extremely important to safety. In addition to conventional deterministic calculations to confirm RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses are interesting because some key variables, albeit conventionally taken at conservative values, can be modeled more accurately through statistical variability. One variable which significantly affects RPV structural integrity assessment is cleavage fracture initiation toughness. The reference fracture toughness method currently in use at EDF is the RCCM and ASME Code lower-bound KIC based on the indexing parameter RTNDT. However, in order to quantify the toughness scatter for probabilistic analyses, the master curve method is being analyzed at present. Furthermore, the master curve method is a direct means of evaluating fracture toughness based on KJC data. In the framework of the master curve investigation undertaken by EDF, this article deals with the following two statistical items: building a master curve from an extract of a fracture toughness dataset (from the European project “Unified Reference Fracture Toughness Design curves for RPV Steels”) and controlling statistical uncertainty for both mono-temperature and multi-temperature tests. Concerning the first point, master curve temperature dependence is empirical in nature. To determine the “original” master curve, Wallin postulated that a unified description of fracture toughness temperature dependence for ferritic steels is possible, and used a large number of data corresponding to nuclear-grade pressure vessel steels and welds. Our working hypothesis is that some ferritic steels may behave in slightly different ways. Therefore we focused exclusively on the basic french reactor vessel metal of types A508 Class 3 and A 533 grade B Class 1, taking the sampling level and direction into account as well as the test specimen type. As for the second point, the emphasis is placed on the uncertainties in applying the master curve approach. For a toughness dataset based on different specimens of a single product, application of the master curve methodology requires the statistical estimation of one parameter: the reference temperature T0. Because of the limited number of specimens, estimation of this temperature is uncertain. The ASTM standard provides a rough evaluation of this statistical uncertainty through an approximate confidence interval. In this paper, a thorough study is carried out to build more meaningful confidence intervals (for both mono-temperature and multi-temperature tests). These results ensure better control over uncertainty, and allow rigorous analysis of the impact of its influencing factors: the number of specimens and the temperatures at which they have been tested.

Fracture Toughness Transferability Study Between the Master Curve Method and a Pressure Vessel Nozzle Using Local Approach

2003 ◽  
Author(s):  
Anssi Laukkanen ◽  
Pekka Nevasmaa ◽  
Heikki Keina¨nen ◽  
Kim Wallin
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Reference Temperature ◽  
Cleavage Crack ◽  
Local Approach ◽  
Master Curve Method ◽  
Curve Method ◽  
Constitutive Parameters

Local approach methods are to greater extent used in structural integrity evaluation, in particular with respect to initiation of an unstable cleavage crack. However, local approach methods have had a tendency to be considered as methodologies with ‘qualitative’ potential, rather than quantitative usage in realistic analyses where lengthy and in some cases ambiguous calibration of local approach parameters is not feasible. As such, studies need to be conducted to illustrate the usability of local approach methods in structural integrity analyses and improve upon the transferability of their intrinsic, material like, constitutive parameters. Improvements of this kind can be attained by constructing improved models utilizing state of the art numerical simulation methods and presenting consistent calibration methodologies for the constitutive parameters. The current study investigates the performance of a modified Beremin model by comparing integrity evaluation results of the local approach model to those attained by using the constraint corrected Master Curve methodology. Current investigation applies the Master Curve method in conjunction with the T-stress correction of the reference temperature and a modified Beremin model to an assessment of a three-dimensional pressure vessel nozzle in a spherical vessel end. The material information for the study is extracted from the ‘Euro-Curve’ ductile to brittle transition region fracture toughness round robin test program. The experimental results are used to determine the Master Curve reference temperature and calibrate local approach parameters. The values are then used to determine the cumulative failure probability of cleavage crack initiation in the model structure. The results illustrate that the Master Curve results with the constraint correction are to some extent more conservative than the results attained using local approach. The used methodologies support each other and indicate that with the applied local approach and Master Curve procedures reliable estimates of structural integrity can be attained for complex material behavior and structural geometries.

Fracture Toughness Evaluation of Reactor Pressure Vessel Steels by Master Curve Method Using Miniature Compact Tension Specimens

2015 ◽  
Vol 137 (5) ◽  
Author(s):  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
Takuyo Ohtsu ◽  
Makoto Udagawa ◽  
Jinya Katsuyama ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Loading Rate ◽  
Compact Tension ◽  
Master Curve Method ◽  
Toughness Evaluation ◽  
Curve Method ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure

We conducted fracture toughness testing on five types of commercially manufactured steel with different ductile-to-brittle transition temperatures. This was performed using specimens of different sizes and shapes, including the precracked Charpy-type (PCCv), 0.4T-CT, 1T-CT, and miniature compact tension specimens (0.16T-CT). Our objective was to investigate the applicability of 0.16T-CT specimens to fracture toughness evaluation by the master curve method for reactor pressure vessel (RPV) steels. The reference temperature (To) values determined from the 0.16T-CT specimens were overall in good agreement with those determined from the 1T-CT specimens. The scatter of the 1T-equivalent fracture toughness values obtained from the 0.16T-CT specimens was equivalent to that obtained from the other larger specimens. Furthermore, we examined the loading rate effect on To for the 0.16T-CT specimens within the quasi-static loading range prescribed by ASTM E1921. The higher loading rate gave rise to a slightly higher To, and this dependency was almost the same for the larger specimens. We suggested an optimum test temperature on the basis of the Charpy transition temperature for determining To using the 0.16T-CT specimens.

Re-Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2015 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hu Hui ◽  
Hui Li ◽  
Fuzhen Xuan
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Master Curve Method ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RTPTS values were determined by the tangent approach. In the case that the RTNDT at or beyond the RPV design life may exceed the RTPTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RTPTS values are increased to varied extents for the different loading transients.

Applicability of Master Curve Method to Japanese Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
Naoki Miura ◽  
Naoki Soneda ◽  
Taku Arai ◽  
Kenji Dohi
Keyword(s):  
Fracture Toughness ◽  
Test Temperature ◽  
Pressure Vessel ◽  
Master Curve ◽  
Loading Rate ◽  
Specimen Size ◽  
Master Curves ◽  
Master Curve Method ◽  
Curve Method ◽  
Reactor Pressure

The Master Curve method has been proposed and recognized worldwide as an alternative approach to evaluate fracture toughness of reactor pressure vessel (RPV) steels in brittle-to-ductile transition temperature range. This method theoretically provides the confidence levels of fracture toughness in consideration of the statistical distribution, which is an inherent property of fracture toughness. In this study, a series of fracture toughness tests was conducted for typical Japanese RPV steels, SFVQ1A and SQV2A, to identify the effects of test temperature, specimen size, and loading rate, and the applicability of the Master Curve method was experimentally validated. The differences in test temperature and specimen size did not affect master curves. In contrast, increasing loading rate significantly shifted master curves to higher temperatures. The lower bound curve based on the master curve could conservatively envelop all of the experimental fracture toughness data. The present rule, in which the lower limit of fracture toughness is indirectly determined by Charpy impact test results, can be too conservative, while the application of the Master Curve method may significantly reduce the conservativity of the allowable level of fracture toughness.

Application of Master Curve Data for Reactor Vessel Steels

2003 ◽  
Author(s):  
William L. Server ◽  
Timothy J. Griesbach ◽  
Stan T. Rosinski
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Material Behavior ◽  
Data Sets ◽  
Master Curve Method ◽  
Curve Method ◽  
Curve Fracture ◽  
Reactor Pressure

The Master Curve method has been developed to determine fracture toughness of a specific material in the brittle-to-ductile transition range. This method is technically more descriptive of actual material behavior and accounts for the statistical nature of fracture toughness properties as an alternative to the current ASME Code reference toughness curves. The Master Curve method uses a single temperature, To, as an index of the Master Curve fracture toughness transition temperature. This method has been successfully applied to numerous fracture toughness data sets of pressure vessel steels contained in the Master Curve database, including the beltline materials for the Kewaunee reactor pressure vessel. The database currently contains over 5,500 toughness data records for vessel weld, plate and forging materials, and it is currently being updated to include more recent fracture toughness data. Application of Master Curve fracture toughness data to reactor pressure vessel (RPV) integrity evaluations requires some assumptions relative to the degree of constraint in the fracture toughness test specimens versus the actual assumed RPV flaw. An excessive degree of conservatism can be introduced if the constraint levels are substantially different. In performing a Master Curve evaluation, the analysis may be restricted by the type of fracture toughness data available. Any excess conservatism should be appropriately considered when the overall safety margin is applied. For example, the precracked Charpy three-point bend specimen actually has some advantages over the compact tension specimen when the application involves a shallow surface flaw in a RPV wall. This paper analyzes some key fracture toughness results from several weld data sets containing both unirradiated and irradiated data to evaluate constraint effects in fracture toughness and pre-cracked Charpy specimens. The evaluated To values were compared to determine if there is any difference in bias from specimen geometry between the unirradiated and irradiated data.

Further Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2016 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hui Hu ◽  
Hui Li
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Master Curve Method ◽  
Pressurized Thermal Shock ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. An earlier work on the PTS analysis of the Chinese Qinshan 300-MWe RPV was performed with the single parameter fracture mechanics method by Shanghai nuclear engineering research and design institute (SNERDI). The integrity analysis of this RPV under PTS was re-evaluated using the Master Curve method later in the paper PVP2015-45577[1]. The objective of this paper is to expand on the previous work, covering more crack geometries and transients to discuss the differences in the use of Master curve based and single parameter linear elastic fracture mechanics based method for PTS analysis. Attempts are made to consider additional size adjustment to the long crack front, which yields more reasonable maximum allowable transition temperature.

Correlations Between Charpy V-Notch Impact Energy and Fracture Toughness of Nuclear Reactor Pressure Vessel (RPV) Steels

2015 ◽  
Author(s):  
Meifang Yu ◽  
Zhen Luo ◽  
Y. J. Chao
Keyword(s):  
Fracture Toughness ◽  
Transition Region ◽  
Test Data ◽  
Impact Energy ◽  
Master Curve ◽  
Reference Temperature ◽  
Master Curve Method ◽  
Curve Method ◽  
Reactor Pressure ◽  
Actual Test

Both Charpy V-notch (CVN) impact energy and fracture toughness are parameters reflecting toughness of the material. Charpy tests are however easy to perform compared to standard fracture toughness tests, especially when the material is irradiated and quantity is limited. Correlations between the two parameters are therefore of great significance, especially for reactor pressure vessel (RPV) structural integrity assessment. In this paper, correlations between CVN impact energy and fracture toughness of three commonly used RPV steels, namely Chinese A508-3 steel, USA A533B steel, Euro 20MnMoNi55 steel, are investigated with two methods. One method applies a direct conversion using empirical formulas and the other adopts the Master Curve method. It is found that when the empirical formula is used, the difference between the predicted fracture toughness (from the CVN impact energy) and actual test data is relatively small in upper shelf, lower shelf and the bottom of transition region, while relatively large in other parts of the transition region. When the Master Curve method is adopted, whether the reference temperature T0 is estimated through temperature at 28J or 41J CVN impact energy, the predicted fracture toughness values of the three steels are consistent with actual test data. The reference temperature T0 is also estimated through the IGC-parameter correlation and through a combination of empirical formula and multi-temperature method. Both procedures show excellent agreement with test results. The mean value of T0 estimated from T28J, T41J, IGC-parameters and the combination method is denoted by TQ-ave and is then used as the final reference temperature T0 for the Master Curve determination. Accuracy of TQ-ave (and therefore the Master Curve method) is demonstrated by comparison with actual test data of the three RPV steels. It is concluded that Master Curve method provides a reliable procedure for predicting fracture toughness in the transition region utilizing limited CVN impact energy data from open literature.

Fracture Toughness Evaluation of Reactor Pressure Vessel Steels by Master Curve Method Using Mini-CT Specimens

2013 ◽  
Author(s):  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
Takuyo Ohtsu ◽  
Makoto Udagawa ◽  
Jinya Katsuyama ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Loading Rate ◽  
Rate Effect ◽  
Master Curve Method ◽  
Toughness Evaluation ◽  
Curve Method ◽  
Temperature Reference ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure

To examine the applicability of Mini-CT (0.16T-CT) specimens to fracture toughness evaluation by Master Curve method, we conducted fracture toughness tests using specimens with different size and shapes such as pre-cracked Charpy-type, 0.4T-CT and 1T-CT in addition to 0.16T-CT specimens for commercially manufactured five kinds of SA533B Cl.1 steels with different ductile-to-brittle transition temperature. Reference temperature To determined by 0.16T-CT specimens were approximately equal to those of 1T-CT specimens for all materials. The Weibull slope of 0.16T-CT specimens was similar to those of other larger specimens. We also examined a loading rate effect on To of 0.16T-CT specimens within the quasi-static loading range prescribed by ASTM E1921. There was no loading rate effect peculiar to 0.16T-CT specimens, while the higher loading rate gave rise to slightly higher To.

Addressing NRC Concerns Regarding Proposed CC N-830: Direct Use of Fracture Toughness for Flaw Evaluations of Pressure Boundary Materials in Class 1 Ferritic Steel Components

2019 ◽  
Author(s):  
Marjorie Erickson ◽  
Mark Kirk
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Pressure Vessel ◽  
Master Curve ◽  
Crack Arrest ◽  
Nuclear Regulatory Commission ◽  
Ferritic Steels ◽  
Class 1 ◽  
Regulatory Commission ◽  
Linkage Models

Abstract Section XI of the ASME Boiler and Pressure Vessel Code provides KIc and KIa fracture toughness models for ferritic steels. These models are based on linear elastic fracture mechanics methods and were initially developed in the 1970s; they remain largely unchanged since that time. Recently, a modification to Code Case (CC) N-830 has been proposed to provide alternative fracture toughness models for use in the flaw evaluation methodologies of ASME Section XI Nonmandatory Appendices A and K. The integrated models contained in proposed Code Case revision predict the mean trends and scatter of the fracture toughness behavior of ferritic steels throughout the temperature range from the lower shelf to the upper shelf. These models include the transition fracture toughness Master Curve and crack arrest master curve approaches that describe the temperature dependence and scatter in KJc and KIa, respectively in the lower transition temperature region. Also included is a model describing the temperature dependence and scatter of JIc on the upper shelf. Finally, linkage models quantify the inter-relationships between these toughness metrics and how they change due to the irradiation-induced hardening. Together, these models describe the temperature dependence and scatter of fracture toughness initiation and arrest behavior for all ferritic reactor pressure vessel (RPV) steels from lower shelf through transition to the upper shelf, all indexed to a single parameter: T0. In late 2017 the Electric Power Research Institute (EPRI) published a report, MRP-418, providing the technical basis for these revisions to CC N-830. Nuclear Regulatory Commission (NRC) staff review of the revised Code Case and MRP-418 resulted in substantive questions regarding validation and range of applicability of the various toughness models. An on-going effort addresses these concerns, and a revision to MRP-418 scheduled for publication later in 2019 will summarize that work. This paper describes the efforts of the WGFE CC-N-830 group to respond to the NRC’s comments, and summarizes responses to some of the comments.

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