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.

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.

The Interrelationships of KIa, KIc, and JIc, and the Implications of These Relationships on Use of Fracture Models Over the Ranges of Hardening Observed in Ferritic Steels

2006 ◽  
Author(s):  
Marjorie EricksonKirk ◽  
Mark EricksonKirk ◽  
Tim Williams
Keyword(s):  
Fracture Toughness ◽  
Crack Initiation ◽  
Master Curve ◽  
Crack Arrest ◽  
Nuclear Industry ◽  
Ferritic Steels ◽  
Initiation Toughness ◽  
Cleavage Crack ◽  
The Mean ◽  
Arrest Toughness

Models to predict the fracture and arrest behavior of ferritic steels, particularly those in use in the nuclear industry, have long been under development. The current, most widely accepted model of fracture toughness behavior is the ASTM E1921-02 “Master Curve” that is used to predict the variation of the mean cleavage fracture toughness with temperature in the transition temperature region as well as predicting the scatter of data about the mean at any given temperature. Recently, models describing the variation of arrest fracture toughness and of ductile initiation toughness with temperature have also been proposed. A study has been conducted with the goal of assessing how the scatter in cleavage initiation toughness may vary with temperature and level of irradiation embrittlement, which utilizes the crack arrest and ductile crack initiation models to redefine limits of applicability of the Master Curve-assumed Weibull distribution by developing empirically-derived interrelationships between the three models. These relationships are expected as all three parameters, KIc, KIa, and JIc, are controlled by the flow behavior of the material. There is a physical basis for viewing the crack arrest toughness as an absolute lower bound to the distribution of crack initiation toughness values for a fixed material condition and temperature. This physically based relationship, borne of the fact that both cleavage crack initiation toughness and cleavage crack arrest toughness are controlled by dislocation mobility, has brought about the suggestion that crack arrest toughness could be used to modify the lower tails of the crack initiation fracture toughness distribution. Using both empirical evidence and a hardening model proposed by Natishan and Wagenhofer, we investigate the relationship between initiation and arrest toughness and the implications on use of toughness models.

Embrittlement Trend Curves for Notch and Fracture Toughness of 15Kh2MFAA Type Steel

2017 ◽  
Author(s):  
Milan Brumovsky ◽  
Radim Kopřiva ◽  
Miloš Kytka
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Pressure Vessel ◽  
Master Curve ◽  
Charpy Impact ◽  
Type Steel ◽  
Impact Tests ◽  
Surveillance Programs ◽  
Vessel Integrity ◽  
Reactor Pressure

Reactor pressure vessel integrity and lifetime evaluation is based on the use of fracture mechanics apparatus but most of the material vessel material data and their degradation during operation are based on results from Charpy V-notch impact tests. Then, empirical correlations between transition shift of temperature dependence of notch toughness and fracture toughness are applied. Elaboration of „Master Curve“ approach for fracture toughness experimental data analysis allows to use fracture toughness data directly to the reactor pressure vessel integrity evaluation. Wider use of this approach is limited by the lack of appropriate database from surveillance specimen test data, as mostly only Charpy impact specimen are included into the Surveillance specimen programs. Fortunately, all WWER Surveillance programs contain also fracture toughness specimens, either pre-cracked Charpy size or CT-0.5. Thus, database of fracture toughness data from Surveillance programs of WWER-440/V-213C type reactor pressure vessels, operated in the Czech Republic, Slovakia and Hungary and manufactured only by one manufacturer - SKODA JS, was collated and analyzed. These vessels were manufactured from 15Kh2MFAA type steel and appropriate weld metal, both of Cr-Mo-V type with low content of detrimental impurities — P and Cu. Analysis of the data in fluence interval up to 6×1024 m−2 (with neutron energies En larger than 0.5 MeV) show that transition temperature shifts in fracture toughness temperature dependence are higher than for Charpy impact tests. Several formulae have been applied for fitting these shift dependencies with chemical composition of materials and finally new Embrittlement Trend Curves for Charpy shifts have been corrected. Additionally, new Embrittlement Trend Curves for fracture toughness shifts based on “Master Curve” approach have been also proposed. Both trends are using simple power law on fluence with exponents around 0.6 and depend on phosphorus and copper contents even though effect of other elements has been also checked.

Remarks to the Upper Shelf Master Curve Concept

2008 ◽  
Author(s):  
Jiri Novak
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Ductile Fracture ◽  
Master Curve ◽  
Critical Strain ◽  
Peierls Stress ◽  
Transition Curve ◽  
Dislocation Slip ◽  
Ferritic Steels ◽  
Deformation Work

Recently, several works appeared in which temperature dependence of ductile fracture toughness of ferritic steels on the upper shelf of brittle-ductile transition curve was analyzed and Upper Shelf Master Curve concept was formulated. Generally, fracture toughness at different temperatures characterized by JIc or dJ/da should be proportional to the deformation work of unit volume evaluated from zero to the critical strain for ductile fracture. As in many other cases, critical strain for ductile fracture initiation may be identified with critical strain for initiation of shear bands, calculated for hyperelastic material with the corresponding stress-strain curve. This concept is successful, among others, in determination of the temperature dependence of fracture toughness of ferritic steels on the upper shelf. Two most important physical mechanisms controlling temperature dependence of constitutive behaviour of ferritic steels in the corresponding temperature range, hence the temperature dependence of both deformation work to initiation of ductile fracture and fracture toughness, are friction resistance to dislocation slip (Peierls stress) and dynamic recovery (dislocation annihilation). Predicted Upper Shelf Master Curve shape based on temperature dependence of constitutive parameters of different ferritic steels corresponds well to the published data.

Effect of Loading Rate on Fracture Toughness of Pressure Vessel Steels

2000 ◽  
Vol 122 (2) ◽  
pp. 125-129 ◽  
Author(s):  
K. K. Yoon ◽  
W. A. Van Der Sluys ◽  
K. Hour
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Loading Rate ◽  
Dynamic Fracture Toughness ◽  
Reference Temperature ◽  
Ferritic Steels ◽  
Transition Range ◽  
High Loading Rate ◽  
Pressure Vessel Steels

The master curve method has recently been developed to determine fracture toughness in the brittle-to-ductile transition range. This method was successfully applied to numerous fracture toughness data sets of pressure vessel steels. Joyce (Joyce, J. A., 1997, “On the Utilization of High Rate Charpy Test Results and the Master Curve to Obtain Accurate Lower Bound Toughness Predictions in the Ductile-to-Brittle Transition, Small Specimen Test Techniques,” Small Specimens Test Technique, ASTM STP 1329, W. R. Corwin, S. T. Rosinski, and E. Van Walle, eds., ASTM, West Conshohocken, PA) applied this method to high loading rate fracture toughness data for SA-515 steel and showed the applicability of this approach to dynamic fracture toughness data. In order to investigate the shift in fracture toughness from static to dynamic data, B&W Owners Group tested five weld materials typically used in reactor vessel fabrication in both static and dynamic loading. The results were analyzed using ASTM Standard E 1921 (ASTM, 1998, Standard E 1921-97, “Standard Test Method for the Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range,” 1998 Annual Book of ASTM Standards, 03.01, American Society for Testing and Materials, West Conshohocken, PA). This paper presents the data and the resulting reference temperature shifts in the master curves from static to high loading rate fracture toughness data. This shift in the toughness curve with the loading rate selected in this test program and from the literature is compared with the shift between KIc and KIa curves in ASME Boiler and Pressure Vessel Code. In addition, data from the B&W Owners Group test of IAEA JRQ material and dynamic fracture toughness data from the Pressure Vessel Research Council (PVRC) database (Van Der Sluys, W. A., Yoon, K. K., Killian, D. E., and Hall, J. B., 1998, “Fracture Toughness of Ferritic Steels and ASTM Reference Temperature T0,” BAW-2318, Framatome Technologies. Lynchburg, VA) are also presented. It is concluded that the master curve shift due to loading rate can be addressed with the shift between the current ASME Code KIc and KIa curves. [S0094-9930(00)01302-0]

Alternative Reference Temperature Based on Master Curve Approach in Japanese Reactor Pressure Vessel Steel

2013 ◽  
Author(s):  
Takatoshi Hirota ◽  
Takashi Hirano ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Reference Temperature ◽  
Ferritic Steels ◽  
Pressure Vessel Steel ◽  
Reactor Pressure

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011, “Test Method for Determination of Reference Temperature, To, of Ferritic Steels” was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature, To can be determined in accordance with this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels including base metals and weld metals, the method for determination of the alternative reference temperature RTTo based on Master Curve reference temperature To was statistically examined, so that RTTo has an equivalent safety margin to the conventional RTNDT. Through the statistical treatment, the alternative reference temperature RTTo was proposed as the following equation; RTTo = To + CMC + 2σTo. This method is applicable to the Japan Electric Association Code JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” as an option item.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

scholarly journals Use of Mini-CT Specimens for Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Before and After Irradiation

2018 ◽  
Author(s):  
Mikhail A. Sokolov
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Collaborative Program ◽  
Before And After ◽  
Minimum Number ◽  
Real Practice ◽  
Reactor Pressure ◽  
Good Agreement

Mini-CT specimens are becoming a highly popular geometry for use in reactor pressure vessel (RPV) community for direct measurement of fracture toughness in the transition region using the Master Curve methodology. In the present study, Mini-CT specimens were machined from previously tested Charpy specimens of the Midland low upper-shelf Linde 80 weld in both, unirradiated and irradiated conditions. The irradiated specimens have been characterized as part of a joint ORNL-EPRI-CRIEPI collaborative program. The Linde 80 weld was selected because it has been extensively characterized in the irradiated condition by conventional specimens, and because of the need to validate application of Mini-CT specimens for low upper-shelf materials — a more likely case for some irradiated materials of older generation RPVs. It is shown that the fracture toughness reference temperatures, To, derived from these Mini-CT specimens are in good agreement with To values previously recorded for this material in the unirradiated and irradiated conditions. However, this study indicates that in real practice it is highly advisable to use a much larger number of specimens than the minimum number prescribed in ASTM E1921.

Master Curve and J-R Fracture Toughness of SA508/SA533-B-1 Weld and HAZ

2013 ◽  
Author(s):  
Randy K. Nanstad ◽  
Xiang Chen ◽  
Mikhail A. Sokolov ◽  
Barry H. Rabin ◽  
Ying Yang
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Master Curve ◽  
Potential Drop ◽  
Material Behavior ◽  
Normalization Procedure ◽  
Threshold Temperatures ◽  
Class 1 ◽  
Large Heat ◽  
Grade 3

A large heat of low-alloy steel that met both specifications for SA508 Grade 3 Class1 forging steel and SA533 Type B Class 1 plate steel (A508/A533) was procured and used to fabricate a submerged-arc weldment for potential application in high temperature gas-cooled reactors. Compact specimens, 1TC(T), were machined from the weld metal and from the heat-affected-zone (HAZ) of the weldment. Tests of both materials were performed to obtain the fracture toughness reference temperature, To, using the Master Curve procedure of ASTM E-1921, and J-R curves to evaluate material behavior at various threshold temperatures in Code Case N-499-2 (2001) of the ASME Boiler and Pressure Vessel Code. Tests were performed at various temperatures up to 593°C. Unloading compliance was the primary technique used, although dc-potential drop was also monitored during the tests, and the normalization procedure of E1820 was used to compare the results from each procedure. Moreover, many tests at the highest temperatures were performed with no unloading and the normalization procedure provided in E1820 was used to analyze the load-displacement measurements. The fracture toughness for the HAZ is superior to that of the weld metal both in terms of transition temperature and ductile fracture toughness.

Fracture Toughness Properties of SA540 Steels for Nuclear Bolting Applications

1977 ◽  
Vol 99 (3) ◽  
pp. 419-426
Author(s):  
R. R. Seeley ◽  
W. A. Van Der Sluys ◽  
A. L. Lowe
Keyword(s):  
Fracture Toughness ◽  
Impact Energy ◽  
Nuclear Reactor ◽  
Charpy Impact ◽  
Nuclear Regulatory Commission ◽  
Charpy Impact Energy ◽  
Energy Correlation ◽  
Impact Properties ◽  
Minimum Requirements ◽  
Regulatory Commission

Large bolts manufactured from SA540 Grades B23 and B24 are used on nuclear reactor vessels and require certain minimum mechanical properties. A minimum fracture toughness of 125 ksi in. (137 MPa m) at maximum operating stresses is required by the Nuclear Regulatory Commission for these bolts. This minimum toughness property was determined by a stress analysis of a bolt. Minimum required Charpy impact properties were calculated by a fracture toughness-Charpy impact energy correlation and the minimum calculated fracture toughness. The fracture toughness, yield strength and Charpy V notch impact properties were determined for five commercial heats of SA540 steels. Correlations between the fracture toughness and Charpy impact properties of these materials were evaluated. The toughness-impact energy correlation used to set the minimum required Charpy impact properties was found to be unduly conservative, and a different correlation of these properties is suggested. The SA540 steels investigated exhibited fracture toughness properties in excess of the NRC minimum requirements.

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