Shengjun Yin

Effect of Shallow Flaws and Biaxial Loading on Transition Temperatures Using a Weibull Stress Model

The aim of the VOCALIST analytical work of the European Commission Project VOCALIST [1] is to analyze the different constraint conditions of tested specimens and to develop a method to describe and predict the constraint dependent shift of the transition temperature T0 for the analyzed specimens. This document describes the analytical work performed within VOCALIST for the biaxial loaded cruciform bend specimens at Oak Ridge National Laboratory (ORNL) USA and FRAMATOME-ANP (FANP) Germany.

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

Application of a Weibull Stress Model to Predict the Failure of Surface and Embedded Cracks in Large Scale Beams Made of Clad and Unclad RPV Steel

Within the European Network NESC, the project NESC IV deals with constraint effects of cracks in large scale beam specimens, loaded by uni- or biaxial bending moments and containing surface or embedded cracks. The specimens are fabricated from original US RPV material, being cladded or cladding is removed. All large scale tests have been conducted at ORNL outside the NESC IV project. The outcome and the analyses of these uncladded and cladded beams containing the surface or embedded cracks are shown. By means of the finite element method, local approach methods and the Weibull stress models the specimens are analysed at the test temperatures and the probability of failure is calculated, taking into account constraint effects. For the case of the embedded cracks it turned out that the failure moment of the uncladded beam is 5% lower than the one of the cladded beam. Both crack fronts of the embedded crack are supposed to fail at the same failure moment. The results of the analysis of the cladded beam showed that the upper crack front nearer to the surface fails prior to the lower crack front, which is located deeper in the specimen (the failure moment is 5% lower). The numerical results agree very well with the experiments. The experimental failure moments could be well predicted and the failure scenario (which crack front fails first) could be determined. A theoretical shift in the transition temperature T0 due to constraint effects could be defined for both crack fronts.Copyright

Derivation of the New Pressurized Thermal Shock Screening Criteria

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.Copyright

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

GRIZZLY/FAVOR Interface Project Report

As part of the Light Water Reactor Sustainability (LWRS) Program, the objective of the GRIZZLY/FAVOR Interface project is to create the capability to apply GRIZZLY 3-D finite element (thermal and stress) analysis results as input to FAVOR probabilistic fracture mechanics (PFM) analyses. The one benefit of FAVOR to Grizzly is the PROBABILISTIC capability. This document describes the implementation of the GRIZZLY/FAVOR Interface, the preliminary verification and tests results and a user guide that provides detailed step-by-step instructions to run the program.

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

ORNL Pre-Test Analyses of a Large-Scale Experiment in STYLE

Oak Ridge National Laboratory (ORNL) is conducting a series of numerical analyses to simulate a large scale mock-up experiment planned within the European Network for S tructural Int egrity for L ifetime Manage ment – non-RPV Components (STYLE). STYLE is a European cooperative effort to assess the structural integrity of (non-reactor pressure vessel) reactor coolant pressure boundary components relevant to ageing and life-time management and to integrate the knowledge created in the project into mainstream nuclear industry assessment codes. ORNL contributes “work-in-kind” support to STYLE Work Package 2 (Numerical Analysis/Advanced Tools) and Work Package 3 (Engineering Assessment Methods/LBB Analyses). This paper summarizes the current status of ORNL analyses of the STYLE Mock-Up3 large-scale experiment to simulate and evaluate crack growth in a cladded ferritic pipe. The analyses are being performed in two parts. In the first part, advanced fracture mechanics models are being developed and performed to evaluate several experiment designs taking into account the capabilities of the test facility while satisfying the test objectives. Then these advanced fracture mechanics models will be utilized to simulate the crack growth in the large scale mock-up test. For the second part, the recently developed ORNL SIAM-PFM open-source, cross-platform, probabilistic computational tool will be used to generate an alternative assessment for comparison with the advanced fracture mechanics model results. The SIAM-PFM probabilistic analysis of the Mock-Up3 experiment will utilize fracture modules that are installed into a general probabilistic framework. The probabilistic results of the Mock-Up3 experiment obtained from SIAM-PFM will be compared to those results generated using the deterministic 3D nonlinear 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

Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.© 2010 ASME

A Generalization of the FAVOR Code to Include BWR Geometries and Heat-Up Transients

The FAVOR computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding, has been and continues to be extensively applied by analysts from the nuclear industry and regulators at the NRC to apply established fracture mechanics and risk-informed methodologies to assess / update regulations designed to insure that the structural integrity of aging and increasingly radiation-embrittled nuclear reactor pressure vessels (RPVs) is maintained throughout the life of the reactor. Earlier versions of FAVOR were primarily developed to perform probabilistic fracture mechanics (PFM) analyses of RPVs subjected to thermal hydraulic transients associated with accidental pressurized thermal shock (PTS) conditions and therefore were limited to modeling internal surface breaking flaws and / or embedded flaws near the RPV internal (wetted) surface. For cool-down transients, these flaws are particularly vulnerable, because at the inner surface the temperature is at its minimum and the tensile stress and radiation-induced embrittlement are at their maximum. Tensile stresses tend to open existing cracks located on or near the internal surface of a reactor pressure vessel (RPV). These earlier versions of FAVOR did not have the capability to model external-surface breaking flaws and / or embedded flaws near the RPV outer surface which are the primary flaws of concern for heat-up transients, such as those associated with reactor start-up. Furthermore, earlier versions of FAVOR were limited to the calculation of applied stress intensity factors (applied KI) of internal surface breaking flaws in RPVs with an internal radius to wall thickness (Ri / t) ratio of approximately 10:1, characteristic of pressurized water reactors (PWRs). This limitation is because the stress intensity factor-influence coefficients (SIFICs), applied by FAVOR to calculate applied KI for surface breaking flaws, were applicable only to internal-surface breaking flaws in RPV geometries characteristics of PWRs. Most boiling water reactors (BWRs) have an (Ri / t) ratio of approximately 20:1, although a few BWRs in the United States have an (Ri / t) ratio of approximately 15. Work has recently been performed at ORNL to generalize the capabilities of the next version of FAVOR, and its successors, such that it will have the capability to perform deterministic and PFM analyses of cool-down and heat-up transients on all domestic commercial PWR and BWR RPV geometries. This paper provides an overview of this generalization of the FAVOR fracture mechanics computer code.Copyright