Materials Aging Degradation of Reactor Vessel Internals: Part III—Structural Evaluation and Local Finite Element Models

2009 ◽  
Author(s):  
Matthew D. Snyder ◽  
Tama´s R. Liszkai ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Structural Analysis ◽  
Constitutive Model ◽  
Reactor Vessel ◽  
Finite Element Models ◽  
Local Models ◽  
Bolted Connections ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) internals components can experience material aging and degradation due to irradiation. The purpose of the functionality analysis is to provide a best-estimate evaluation of the reactor internals core barrel assembly for materials degradation to see if the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis [1] representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox-type (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates of the internals. The computational fluid dynamics domain (CFD) allows evaluation of the internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections. Therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [2] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, to be presented at PVP 2009 [2] describes global structural finite element models. Part III, presented in this paper, presents a description of local models of bolted connections, results, and conclusions.

Materials Aging Degradation of Reactor Vessel Internals: Part II—Structural Evaluation and Global Finite Element Models

2009 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Matthew Snyder ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Structural Analysis ◽  
Constitutive Model ◽  
Reactor Vessel ◽  
Finite Element Models ◽  
Local Models ◽  
Bolted Connections ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), developed PWR Internals Inspection and Evaluation (I&E) Guidelines mainly to support license renewal of U.S. plants [2]. The functionality analysis of reactor internals components and assemblies was one of the tools used to develop these guidelines. The purpose of the functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation and to assess whether the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement [3], 4, [5]. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major analytical calculation, which facilitates the development of operating stress fields within the RV internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections, therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [6] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, detailed in this paper, describes global structural finite element models. Part III, to be also presented at PVP-2009 [7], presents a description of local models of bolted connections, results, and conclusions.

Materials Aging Degradation of Reactor Vessel Internals: Part I — Thermal Hydraulics Evaluation

2008 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Norm S. Yee ◽  
Jim R. Smotrel ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Failure Modes ◽  
Operating Experience ◽  
Material Model ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Essential Information ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), is developing Reactor Internals Inspection and Evaluation (I&E) Guidelines mainly to support U.S. license renewed plants. These guidelines are organized around a framework and strategy, [3] and [4], for managing the effects of aging in PWR internals as shown in Figure 1, dependent on a substantial database of material data and supporting results. The key steps include the following: the development of screening criteria, with susceptibility levels for the eight postulated aging mechanisms relevant to reactor internals and their effects [5]; an initial component screening and categorization step, using the susceptibility levels to identify the relative susceptibility of the components; a functionality assessment of degradation for components and assemblies of components; and finally aging management strategy development combining the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. The purpose of this functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation up to 60 years of operation. The stainless steel material model employed in the calculations is an irradiated material-specific constitutive model for use in a finite element analysis [8]. The material model accounts for the effects of plasticity, irradiation assisted stress corrosion cracking (IASCC), irradiation creep-stress relaxation, void swelling, and embrittlement as a function of temperature and fluence [6] and [7]. The study focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, detailed finite element models were developed capable of representing the complex interactions between the components. The goal of this study is to characterize the potential failure modes, spatial and chronological distribution of component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three major physics fields. Radiation calculations of the core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the RV internals. The three major physics fields and their relations are illustrated in Figure 2. This paper focuses on the CFD/CHT aspects of the overall analysis for the B&W designed RV internals and provides information on the state-of-the-art multi-physics approach employed.

A constitutive model for Fe-based shape memory alloy considering martensitic transformation and plastic sliding coupling: Application to a finite element structural analysis

2012 ◽  
Vol 23 (10) ◽  
pp. 1143-1160 ◽  
Author(s):  
Walid Khalil ◽  
Alain Mikolajczak ◽  
Céline Bouby ◽  
Tarak Ben Zineb
Keyword(s):  
Finite Element ◽  
Phase Transformation ◽  
Structural Analysis ◽  
Shape Memory Alloy ◽  
Constitutive Model ◽  
Shape Memory ◽  
Internal Variables ◽  
Thermomechanical Loading ◽  
Proposed Model ◽  
Numerical Tool

In this article, we propose a finite element numerical tool adapted to a Fe-based shape memory alloy structural analysis, based on a developed constitutive model that describes the effect of phase transformation, plastic sliding, and their interactions on the thermomechanical behavior. This model was derived from an assumed expression of the Gibbs free energy taking into account nonlinear interaction quantities related to inter- and intragranular incompatibilities as well as mechanical and chemical quantities. Two scalar internal variables were considered to describe the phase transformation and plastic sliding effects. The hysteretic and specific behavior patterns of Fe-based shape memory alloy during reverse transformation were studied by assuming a dissipation expression. The proposed model effectively describes the complex thermomechanical loading paths. The numerical tool derived from the implicit resolution of the nonlinear partial derivative constitutive equations was implemented into the Abaqus® finite element code via the User MATerial (UMAT) subroutine. After tests to verify the model for homogeneous and heterogeneous thermomechanical loadings, an example of Fe-based shape memory alloy application was studied, which corresponds to a tightening system made up of fishplates for crane rails. The results we obtained were compared to experimental ones.

Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

scholarly journals Finite Element Modeling of K-Monel Bolts under Static Loading and Dynamic Shock Loading

2013 ◽  
Vol 20 (3) ◽  
pp. 575-589 ◽  
Author(s):  
Kevin Behan ◽  
Emily Guzas ◽  
Jeffrey Milburn ◽  
Stacy Moss
Keyword(s):  
Finite Element ◽  
High Resolution ◽  
Shock Loading ◽  
Future Research ◽  
Finite Element Models ◽  
Computationally Efficient ◽  
Bolted Connections ◽  
Bolted Connection ◽  
Physical Testing ◽  
Static Shear

The Naval Undersea Warfare Center has funded a project to investigate the accuracy of various bolt models used to represent actual shipboard bolted connections within an analytical shock survivability assessment. The ultimate goal within this project is to develop finite element bolt representations that are not only computationally efficient, but also accurate. A significant task within this effort involved the development of highly detailed finite element models of bolted connections under various load configurations. Accordingly, high-resolution bolt models were developed and incorporated into simulations of four bolted connection test arrangements: static shear, static tension, dynamic shear, and dynamic tension. These simulation results are validated against experimental data from physical testing of each configuration. Future research will focus on exploring simplified finite element bolt representations and comparing these against the high-resolution results.

scholarly journals Effects of region and sex on the mechanical properties of the glenohumeral capsule during uniaxial extension

2010 ◽  
Vol 108 (6) ◽  
pp. 1711-1718 ◽  
Author(s):  
Carrie A. Voycheck ◽  
Eric J. Rainis ◽  
Patrick J. McMahon ◽  
Jeffrey A. Weiss ◽  
Richard E. Debski
Keyword(s):  
Finite Element ◽  
Constitutive Model ◽  
Material Properties ◽  
Surgical Repair ◽  
Glenohumeral Joint ◽  
Uniaxial Extension ◽  
Posterior Capsule ◽  
Finite Element Models ◽  
Men And Women ◽  
Glenohumeral Capsule

Surgical repair of the glenohumeral capsule after dislocation ignores regional boundaries of the capsule and is not sex specific. However, each region of the capsule functions to stabilize the joint in different positions, and differences in joint laxity between men and women have been found. The objectives of this research were to determine the effects of region (axillary pouch and posterior capsule) and sex on the material properties of the glenohumeral capsule. Boundary conditions derived from experiments were used to create finite-element models that applied tensile deformations to tissue samples from the capsule. The material coefficients of a hyperelastic constitutive model were determined via inverse finite-element optimization, which minimized the difference between the experimental and finite-element model-predicted load-elongation curve. These coefficients were then used to create stress-stretch curves representing the material properties of the capsule regions for each sex in response to uniaxial extension. For the axillary pouch, the C1(men: 0.28 ± 0.39 MPa and women: 0.23 ± 0.12 MPa) and C2(men: 8.2 ± 4.1 and women: 7.7 ± 3.0) material coefficients differed between men and women by only 0.05 MPa and 0.5, respectively. Similarly, the posterior capsule coefficients differed by 0.15 MPa (male: 0.49 ± 0.26 MPa and female: 0.34 ± 0.20 MPa) and 0.6 (male: 7.8 ± 2.9 and female: 7.2 ± 3.0), respectively. No differences could be detected in the material coefficients between regions or sexes. As a result, surgeons may not need to consider region- and sex-specific surgical repair techniques. Furthermore, finite-element models of the glenohumeral joint may not need region- or sex-specific material coefficients when using this constitutive model.

Influence of Concrete Strength of the Core of the Joints on the Holistic Resistant Behavior of Super High-Rise Structure

2013 ◽  
Vol 351-352 ◽  
pp. 396-400 ◽  
Author(s):  
Zhao Yang ◽  
Xiao Yu ◽  
Yang Zhi Zhong
Keyword(s):  
Finite Element ◽  
Concrete Structure ◽  
Concrete Strength ◽  
Core Area ◽  
Finite Element Models ◽  
Construction Process ◽  
Construction Quality ◽  
High Rise ◽  
The Core ◽  
Design Requirements

In the construction process of the super high-rise concrete structure, it’s easy to be happened that the concrete strength of joints can’t meet design requirements. Some finite element models of a super high-rise building were established by MIDAS in the paper, which were used to analyze the influence of concrete strength of the core of the joints on the holistic resistant behavior. The study may provide the basis for solving the construction quality problems of the core area of joints

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