B. Richard Bass

Shallow Flaws Under Biaxial Loading Conditions—Part I: The Effect of Specimen Size on Fracture Toughness Values Obtained From Large-Scale Cruciform Specimens

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed. This technology is for application to the safety assessment of RPVs containing postulated shallow-surface flaws. It has been shown that relaxation of crack-tip constraint causes shallow-flaw fracture toughness of RPV material to have a higher mean value than that for deep flaws in the lower transition temperature region. Cruciform beam specimens developed at Oak Ridge National Laboratory (ORNL) introduce far-field, out-of-plane biaxial stress components in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock (PTS) loading of an RPV. The biaxial stress component has been shown to increase stress triaxiality (constraint) at the crack tip, and thereby reduce the shallow-flaw fracture toughness enhancement. The cruciform specimen permits controlled application of biaxial loading ratios, resulting in controlled variation of crack-tip constraint. An extensive matrix of intermediate-scale cruciform specimens with a uniform depth surface flaw was previously tested and demonstrated a continued decrease in shallow-flaw fracture toughness with increasing biaxial loading. This paper describes the test results for a series of large-scale cruciform specimens with a uniform depth surface flaw. These specimens were all of the same size with the same depth flaw and were tested at the same temperature and biaxial load ratio (1:1). The configuration is the same as the previous set of intermediate-scale tests, but has been scaled upward in size by 150 percent. These tests demonstrated the effect of biaxial loading and specimen size on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. For specimens tested under full biaxial (1:1) loading at test temperatures in the range of 23°F (-5°C) to 34°F (1°C), toughness was reduced by approximately 15 percent for a 150-percent increase in specimen size. This decrease was slightly greater than the predicted reduction for this increase in specimen size. The size corrections for 1/2T C(T) specimens did not predict the experimentally determined mean toughness values for larger size shallow-flaw specimens tested under biaxial (1:1) loading in the lower transition temperature region.

Analysis of Ductile Crack Growth in Pipe Test in STYLE Project

The Oak Ridge National Laboratory (ORNL) is conducting structural analyses, both deterministic and probabilistic, to simulate a large scale mock-up experiment planned within the European Network for Structural Integrity for Lifetime Management – non-RPV Components (STYLE).The paper summarizes current ORNL analyses of STYLE’s Mock-Up3 experiment to simulate/evaluate ductile crack growth in a cladded ferritic pipe. Deterministic analyses of the large-scale bending test of a ferritic surge pipe, with an internal circumferential crack, are being simulated with a number of local micromechanical approaches, such as Gurson-Tvergaard-Needleman (GTN) model. Both FEACrack [1] and ABAQUS [2] general purpose finite element programs are being used to predict the failure load and the failure mode, i.e. ductile tearing or net-section collapse, as part of the pre-test phase of the project.Companion probabilistic analyses of the experiment are utilizing the ORNL developed open-source Structural Integrity Assessment Modular - Probabilistic Fracture Mechanics (SIAM-PFM) framework. SIAM-PFM contains engineering assessment methodologies such as the tearing instability (J-T analysis) module developed for inner surface cracks under bending load. The driving force J-integral estimations are based on the SC.ENG1 or SC.ENG2 models. The J-A2 methodology is used to transfer (constraint-adjust) J-R curve material data from standard test specimens to the Mock-Up3 experiment configuration. The probabilistic results of the Mock-Up3 experiment obtained from SIAM-PFM will be compared to those generated using the deterministic finite element modeling approach. The objective of the probabilistic analysis is to provide uncertainty bounds that will assist in assessing the more detailed 3D finite-element solutions and to also assess the level of confidence that can be placed in the best-estimate finite-element solutions.Copyright

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

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.© 2003 ASME

Heavy-Section Steel Technology and Irradiation Programs—Retrospective and Prospective Views

In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and, in 2007, the HSST/HSSI Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrated 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels (RPV). This paper presents a summary of results from those programs with a view to future activities.

Effect of Cladding Residual Stress Modeling Technique on Shallow Flaw Stress Intensity Factor in a Reactor Pressure Vessel

Prior probabilistic fracture mechanics (PFM) analysis of reactor pressure vessels (RPVs) subjected to normal cool-down transients has shown that shallow, internal surface-breaking flaws dominate the RPV failure probability. This outcome is caused by the additional crack driving force generated near the clad interface due to the mismatch in coefficient of thermal expansion (CTE) between the cladding and base material, which elevates the thermally induced stresses. The CTE contribution decreases rapidly away from the cladding, making this effect negligible for deeper flaws. The probabilistic fracture mechanics code FAVOR (Fracture Analysis of Vessels, Oak Ridge) uses a stress-free temperature model to account for residual stresses in the RPV wall due to the cladding application process. This paper uses finite element analysis to compare the stresses and stress intensity factor during a cool-down transient for two cases: (1) the existing stress-free temperature model adopted for use in FAVOR, and (2) directly applied RPV residual stresses obtained from empirical measurements made at room temperature. It was found that for a linear elastic fracture mechanics analysis, the application of measured room temperature stresses resulted in a 10% decrease in the peak stress intensity factor during a cool-down transient as compared to the stress-free temperature model.Copyright

The Role of Grain Size in Predicting Cleavage Fracture Toughness of Pressure Vessel Steels

A theoretical, multi-scale model has been developed to predict the fracture toughness of ferritic steels in the ductile-to-brittle fracture mode transition temperature region. The new model is being implemented into the DISlocation-based FRACture (DISFRAC) computer code at the Oak Ridge National Laboratory (ORNL) and will permit fracture safety assessments of ferritic structures with only tensile properties and microstructural information (grain and carbide size) required as input. The theoretical basis of this model provides a means of predicting fracture behavior outside of the ranges of data currently used in deriving empirically-based models and should provide a means of improving the understanding of fracture behavior in the fracture mode transition region. Dislocation distribution equations, derived from dislocation theory developed by Yokobori et al., are combined with modified boundary layer solutions to account for the stress state local to various microstructural features believed to control fracture behavior. Terms are included to account for microcrack initiation in brittle grain boundary particles, propagation of the microcrack into the first ferrite grain and then through subsequent grain boundaries accounting for local tilt and twist grain misorientation across boundaries. This paper summarizes the DISFRAC model and provides the results of a study performed to investigate the role of grain size in microcrack initiation, propagation and the resulting prediction of fracture toughness.Copyright

A Dislocation-Based Cleavage Initiation Model for Pressure Vessel Steels

Efforts are under way to develop a theoretical, multi-scale model for the prediction of fracture toughness of ferritic steels in the ductile-to-brittle transition temperature (DBTT) region that accounts for temperature, irradiation, strain rate, and material condition (chemistry and heat treatment) effects. This new model is intended to address difficulties associated with existing empirically-derived models of the DBTT region that cannot be extrapolated to conditions for which data are unavailable. Dislocation distribution equations, derived from the theories of Yokobori et al., are incorporated to account for the local stress state prior to and following initiation of a microcrack from a second-phase particle. The new model is the basis for the DISlocation-based FRACture (DISFRAC) computer code being developed at the Oak Ridge National Laboratory (ORNL). The purpose of this code is to permit fracture safety assessments of ferritic structures with only tensile properties required as input. The primary motivation for the code is to assist in the prediction of radiation effects on nuclear reactor pressure vessels, in parallel with the EURATOM PERFORM 60 project.

Applicability of (K, T-Stress) Methodology to Analyze RPV Under Thermal-Hydraulic Transients

The (K, T-stress) methodology developed by Gao and Dodds [1] is being utilized to introduce crack front plasticity with constraint effects when plastic deformation occurs in structures, for example, when the Reactor Pressure Vessels (RPVs) are subjected to thermal-hydraulic loadings. One crucial step in this procedure is to quantify combinations of flaw geometries and loading conditions (transient sequences) that illustrate the limits of applicability of the two-parameter (K, T-stress) advanced fracture methodology relevant to integrity analyses of RPVs subjected to normal and emergency operating conditions. Numerical analyses were conducted to determine the limits of applicability of (K, T-stress) advanced fracture technology for RPV under thermal-hydraulic loadings. The numerical results indicate that the (K, T-stress) methodology captures the constraint condition of the RPV with typical embedded flaws under a postulated dominant thermal-hydraulic transient.© 2007 ASME

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

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

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.Copyright