Further Predictions of Fracture Experiments of a Pre-Cracked Pressurised Thermal Shock Disk (PTS-D) Specimen Under Warm Prestressing Loading Cycles

2012 ◽  
Author(s):  
H. Teng ◽  
J. K. Sharples ◽  
P. J. Budden
Keyword(s):  
Fracture Toughness ◽  
Finite Element ◽  
Thermal Shock ◽  
Local Stress ◽  
Superposition Method ◽  
Warm Prestressing ◽  
Disk Specimen ◽  
Pressurized Thermal Shock ◽  
Different Levels ◽  
Good Agreement

Finite element analyses have been performed to investigate the effects of warm prestressing (WPS) of a pre-cracked PTS-D (Pressurized Thermal Shock Disk) specimen. Three basic types of WPS loading cycles were used in the analyses: LUCF (Load-Unload-Cool-Fracture) cycle; LCF (Load-Cool-Fracture) cycle; and LCTF (Load-Cool-Transient-Fracture) cycle. The analyses aimed to predict the fracture toughness enhancements due to WPS using different analysis methods and to make comparisons with the experimental work conducted by the Belgium SCK-CEN organisation under the European NESC VII project. The finite element results were used to derive the enhanced fracture toughness by three different engineering methods: (1) Chell’s displacement superposition method; (2) the local stress matching method; and (3) Wallin’s empirical formula. The enhanced fracture toughness was evaluated at the deepest point of the semi-elliptical crack based on three different levels of as-received fracture toughness of 43.96, 65.94, and 86.23 MPam1/2, which correspond to probabilities of failure of 5%, 50% and 95%, respectively. The predicted fracture loads were compared with the experimental fracture loads for the three WPS loadings cycles. The results show good agreement.

The Effects of Warm Pre-Stressing of a Pre-Cracked Pressurised Thermal Shock Disk Specimen Under a LUCF Loading Cycle

2011 ◽  
Author(s):  
H. Teng ◽  
D. W. Beardsmore ◽  
J. K. Sharples ◽  
P. J. Budden
Keyword(s):  
Fracture Toughness ◽  
Finite Element ◽  
Thermal Shock ◽  
Local Stress ◽  
Superposition Method ◽  
Element Analysis ◽  
Loading Cycle ◽  
Finite Element Software ◽  
Disk Specimen ◽  
Apparent Fracture Toughness

A finite element analysis has been performed to investigate the effects of warm prestressing of a pre-cracked PTS-D (Pressurized Thermal Shock Disk) specimen, for comparison with the experimental work conducted by the Belgium SCK-CEN organisation under the European NESC VII project. The specimen was loaded to a maximum loading at −50 °C, unloaded at the same temperature, cooled down to −150 °C, and then re-loaded to fracture at −150 °C. This is a loading cycle known as a LUCF cycle. The temperature-dependant tensile stress-strain data was used in the model and the finite element software ABAQUS was used in the analysis. The finite element results were used to derive the apparent fracture toughness by three different methods: (1) Chell’s displacement superposition method; (2) the local stress matching method; and (3) Wallin’s empirical formula. The apparent fracture toughness values were derived at the deepest point of the semi-elliptical crack for a 5% un-prestressed fracture toughness of 43.96 MPam1/2 at −150 °C. The detailed results were presented in the paper.

RPV Structure Integrity Assessment With Surface Cracks Under PTS

2017 ◽  
Author(s):  
Wei Lu ◽  
Zheng He
Keyword(s):  
Fracture Toughness ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Stress Distribution ◽  
Intensity Factor ◽  
Thermal Shock ◽  
Surface Crack ◽  
Parameter Analysis ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock

As one of the most critical barrier of pressurized-water reactor, Reactor Pressurized Vessel (RPV) is exposed to high temperature, high pressure and irradiation. During the lifetime of RPV, the core belt material will become brittle under the influence of neutron irradiation. The ductile-brittle transition temperature will increase and upper shelf energy will decrease. Thus the structure integrity evaluation of RPV concerning brittle fracture is one of the most important tasks of RPV lifetime management. The non-LOCA accident of Rancho Seco nuclear power plant in 1978 indicates that the emergent cooling transients the sudden cooling down may accompany with the re-pressurize of main loop. The combination of pressure loads and thermal loads may induce a large tensile stress in RPV internal surface, which is the so called pressurized thermal shock (PTS). Due to the existence of welding cladding on the inner surface of RPV, the discontinuity of stress distribution on the cladding-base interface of RPV wall will make calculation of stress-intensity-factor (SIF) difficult. In present research, a two dimensional axial-symmetrical model is built and Finite Element Method (FEM) is adopted to calculate the transient thermal distribution and stress distribution. The influence function method is adopted to calculate crack SIF. Stress distributions in the base and cladding are decomposed respectively and SIFs are calculated respectively to obtain the crack SIF. ASME method is used to calculate the fracture toughness. Present PTS program is validated by the comparative benchmark calculation (the International Comparative Assessment Study of Pressurized Thermal-Shock in Reactor Pressure Vessels). The calculated SIF from present program lies in the reasonable region of the comparing group results. A LOCA transient is investigated with a semi-elliptical surface crack on the RPV beltline region. The temperature and stress distribution along the vessel wall during the transient are given. The stress intensity factors at the deepest and interface point are given respectively. The integrity of RPV under PTS transient is evaluated by comparing stress intensity factor with fracture toughness. Results indicate that the stress intensity factor will not exceed the fracture toughness of the RPV material. The difference between the stress intensity factor and fracture toughness reach a minimum value at the crack tip temperature 20°C. Present research gives a reliable and efficient program to perform RPV structure integrity assessment with surface crack under PTS, which is suitable for further parameter analysis and probabilistic analysis.

Influence Factors of Fracture Mechanics Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock

2019 ◽  
Vol 795 ◽  
pp. 333-339
Author(s):  
Juan Luo ◽  
Jia Cheng Luo
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Influence Factors ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Mechanics Analysis ◽  
Reactor Pressure

When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

Finite Element Modeling of a Homogeneous Pneumatic Tire Subjected To Footprint Loadings

1985 ◽  
Vol 13 (2) ◽  
pp. 91-110 ◽  
Author(s):  
R. A. Ridha ◽  
K. Satyamurthy ◽  
L. R. Hirschfelt
Keyword(s):  
Finite Element ◽  
Strain Energy Density ◽  
Strain Energy ◽  
Contact Forces ◽  
Experimental Results ◽  
Contact Algorithm ◽  
Footprint Area ◽  
Distribution Of Stress ◽  
Different Levels ◽  
Good Agreement

Abstract In this paper, we use an approach which involves 3D finite elements and a contact algorithm. For a given deflection of the hub, the algorithm computes the nodes that come into contact with a predefined contact plane, and the magnitudes of the contact forces at the nodes. Thus, the algorithm computes the size and shape of the footprint, and the contact forces at the FE nodes. To illustrate the technique, we analyze a homogeneous tire subjected to footprint loading. The computed shapes and sizes of the footprint area at different levels of rim deflection are shown to be in good agreement with experimental results. The computed tire profiles and the load-deflection response of the tire are also in good agreement with experimental results. The computed results include the distribution of stress, strain, and strain energy density within the tire, and the changes in this distribution with applied footprint loadings.

Crack Propagation Calculation for Axial Cracks in Hollow Cylinders Subjected to Thermal Shock

2021 ◽  
Author(s):  
Diego F. Mora M. ◽  
Markus Niffenegger
Keyword(s):  
Fracture Toughness ◽  
Finite Element ◽  
Crack Propagation ◽  
Weight Function ◽  
Thermal Shock ◽  
Initial Crack ◽  
Fe Model ◽  
Simple Method ◽  
Hollow Cylinders ◽  
Fracture Arrest

Abstract The core region of the RPV can be considered a hollow circular cylinder disregarding the geometrical details due to nozzles. This contribution investigates the prediction capabilities for crack initiation, crack growth and arrest by means of a rather simple method based on the closed-weight function formula for the stress intensity factor (SIF) for axial cracks in hollow cylinders subjected to thermal shock. The method is explained together with some illustrative examples for real low allow steel used in nuclear applications. In order to obtain the temperature and stress distribution in the cylinder during the thermal shock, a finite element (FE) model is defined to obtain the uncoupled solution of these two fields needed for the closed-weight function. Since the material exhibits a ductile-brittle transition fracture behavior, the temperature-dependent fracture toughness for initiation and for arrest are described using the ASME model. The solution for the SIF is based on linear elastic fracture mechanics (LEFM) and therefore only elastic material is assumed and the crack can propagate in brittle manner. The crack initiates propagation if the SIF value at the crack tip reaches the fracture toughness (for initiation) and propagates unstably in mode I unless the fracture arrest toughness is reached. The quality of the solution is checked by comparing the obtained solution for a “stationary” crack with the calculated extended finite element method (XFEM) solution for the same loading transient. The results show that for some geometries of the cylinder, the crack stops and in some other cases the crack propagates until the cylinder fails. The combined closed-weight function-initiation-growth-arrest (WFF-IGA) algorithm does not require expensive computational resources and gives fast reliable results. The WFF-IGA method provides a powerful and economical way to predict the crack propagation and arrest of the initial crack. This is an advantage when an optimization of the structure is needed.

Self-tensioning Support Post Design to Control Residual Stress in MEMS Fixed-Fixed Beams

2014 ◽  
Vol 1659 ◽  
pp. 55-61
Author(s):  
Ryan M. Pocratsky ◽  
Maarten P. de Boer
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Finite Element Model ◽  
Analytical Model ◽  
Local Stress ◽  
Element Model ◽  
3D Finite Element ◽  
3D Finite Element Model ◽  
Fixed Beam ◽  
Good Agreement

ABSTRACTFixed-fixed beams are ubiquitous MEMS structures that are integral components for sensors and actuation mechanisms. However, residual stress inherent in surface micromachining can affect the mechanical behavior of fixed-fixed structures, and even can cause buckling. A self-tensioning support post design that utilizes the compressive residual stress of trapped sacrificial oxide to control the stress state passively and locally in a fixed-fixed beam is proposed and detailed. The thickness and length of the trapped oxide affects the amount of stress in the beam. With this design, compression can be reduced or even converted into tension. An analytical model and a 3D finite element model are presented. The analytical model shows relatively good agreement with a 3D finite element model, indicating that it can be used for design purposes. A series of fixed-fixed beams were fabricated to demonstrate that the tensioning support post causes a reduction in buckling amplitude, even pulling the beam into tension. Phase shifting interferometry deflection measurements were used to confirm the trends observed from the models. Controlling residual stress allows longer fixed-fixed beams to be fabricated without buckling, which can improve the performance range of sensors. This technique can also enable local stress control, which is important for sensors.

Thermal Shock Testing of Ceramics for Circuit Substrates

2006 ◽  
Vol 11-12 ◽  
pp. 31-34 ◽  
Author(s):  
Sawao Honda ◽  
Shinobu Hashimoto ◽  
Hideo Awaji
Keyword(s):  
Fracture Toughness ◽  
Thermal Properties ◽  
Thermal Stress ◽  
Thermal Shock ◽  
Alumina Ceramics ◽  
Radiation Heating ◽  
Ambient Temperatures ◽  
Shock Testing ◽  
Experimental Values ◽  
Good Agreement

Thermal shock resistances of commercially available aluminum nitride and alumina ceramics as used for the circuit substrate were evaluated by infrared radiation heating (IRH) technique. Thermal shock fracture toughness, R2c of these materials was estimated experimentally and theoretically using IRH technique at various ambient temperatures. Temperature dependence of thermal properties of the materials was taken into account for the temperature and the thermal stress analysis. Experimental values of thermal shock fracture toughness were in good agreement with the calculated values. Thermal shock fracture toughness decreased with elevated ambient temperature in both ceramics.

In-Service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution Per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

2011 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Michael T. Anderson ◽  
Wallace Norris
Keyword(s):  
Fracture Toughness ◽  
Density Distribution ◽  
Thermal Shock ◽  
Size Distribution ◽  
Pressure Vessel ◽  
Ultrasonic Testing ◽  
Reactor Vessel ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Reactor Pressure

The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking of a reactor pressure vessel (RPV) due to a pressurized thermal shock (PTS) event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” The §50.61a rule, which is optional, requires licensees to analyze the results from periodic volumetric examinations required by the American Society of Mechanical Engineers (ASME) Code. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory has been working on a program to assess the ability of current inservice inspection ultrasonic testing (UT) techniques, as qualified through the ASME Code to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, “Spirit of Appendix VIII reactor vessel examination,” a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

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