Development of a New Specimen to Study Crack Propagation Threshold and Non-Propagation Conditions

2011 ◽  
Author(s):  
Pauline Bouin ◽  
Ste´phane Marie ◽  
Gre´gory Perez
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Complex Loading ◽  
Crack Arrest ◽  
Compact Tension ◽  
Element Analysis ◽  
Loading Ratio ◽  
Nuclear Structures ◽  
Catastrophic Fracture

The extension of nominal service-life of nuclear power plants leads to calculate more involved safety analyses of the nuclear structures for the different scenario of failures. In order to prevent catastrophic fracture, it is important to minimise the initiation of cracks or if considering that a crack can initiate, to determine if the extension of these cracks can stop or lead to the failure of the structure integrity. As a result, this work requires better understanding and characterisation of the crack propagation or non-propagation behaviour. Existing practises to determinate fatigue crack growth rate and fatigue thresholds are mainly based on compact tension (CT) specimen testing. For high loading ratio, the maximum of stress intensity factor (SIF) usually remains constant whereas the minimum of the SIF is increased to reduce the SIF range down to the threshold. On the contrary, for low loading ratio, complex loading control is required to make maximum and minimum SIF decrease. Furthermore, the results of these tests are mainly dependent on the accuracy of the piloting of the test bench, the specimen instrumentation and the force loading cell capacity. This paper presents details of adaptation of a specimen initially developed to study crack arrest problems under cleavage fracture. A bulk compact tension (BCT) specimen has been designed based on a standard CT specimen. Specifically, a reinforced heel at the back of the specimen enables a reduction of the loading at the crack front. In the case of fatigue, this reduces the SIF range while the loading condition remains constant during the test. A SIF calibration for these BCT specimens has also been established to estimate the compliance using finite element analysis with the French code, Cast3M.

Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

Fracture Toughness Characterization of Welds and HAZ at Various Loading Rates Using Differing Specimen Geometries

2011 ◽  
Author(s):  
S. Kalyanam ◽  
M. Uddin ◽  
G. Wilkowski ◽  
D.-J. Shim ◽  
Y. Hioe ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Crack Propagation ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Compact Tension ◽  
Dynamic Strain ◽  
Element Analysis ◽  
Loading Rates ◽  
Fracture Tests

The characterization of the fracture toughness in weld and HAZ materials employed in nuclear power plant (NPP) piping is essential to assess the structural integrity of the piping systems under various loading conditions, when cracks/flaws are present in the weld/HAZ material regions. The current investigation was undertaken to determine the resistance to crack propagation using ASTM E1820 standard based compact tension, 1T-C(T), specimens with initial crack in the weld centerline (WCL) and adapted 1T-C(T) specimens with slanted cracks in the heat-affected-zone (HAZ) materials. The experiments were conducted at loading rates ranging from quasi-static to dynamic rates estimated for seismic loading at normal operating temperature. The fracture tests were also conducted at various loading rates to verify the occurrence of phenomena such as dynamic-strain-aging (DSA) that could potentially cause significant changes in the material toughness, J-R curves of the weld and HAZ materials. The investigation also compared the fracture toughness obtained from 1T-C(T) fracture tests with those obtained from impact loading on single-edge-notch-bend (SENB) specimens in a drop-weight-tear-test (DWTT) machine and others obtained from SEN(B) specimens tested at quasi-static loading rates. The range of crack growth resistance curves obtained from the various fracture tests and specimens were used to develop bounding JIc and J-R curves that are to be employed in computational finite element analysis (FEA) that are used determine crack propagation and stability in the weld/HAZ materials of NPP plant piping.

Development of an Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Hiroshi Okada
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Integral Method ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Element Analysis ◽  
3D Finite Element ◽  
Asme Code ◽  
Virtual Crack Extension Method

When cracks are detected in piping in nuclear power plants during in-service inspections, the crack propagation is usually calculated using approximation formulas of stress intensity factor (SIF) provided in the ASME Code, the JSME Rules or the literature. However, when the crack is detected in complicated-shaped locations in components, finite element analysis (FEA) needs to be used to calculate the SIFs. Accordingly, a method of automatically conducting FEA for crack propagations in nuclear power plants is needed. Therefore, we, the Nuclear Regulation Authority (NRA, Japan) have developed an automatic 3D finite element crack propagation system (CRACK-FEM) for nuclear components. The developed CRACK-FEM uses three methods of SIF calculation: the Virtual Crack Extension Method (VCEM), the Virtual Crack Closure-Integral Method (VCCM) and the Domain Integral Method (DIM). Each method uses different meshes, so users can select a method which uses a suitable mesh for the problem. The software includes a geometry generator to create complicated weld models, and a mesh generator which can deal with interior boundaries formed between different materials. The functions and accuracy of the new software are demonstrated by solving several sample problems involving crack propagation. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

2025 Nuclear Code: The Vision for the Future of ASME Nuclear Codes and Standards

2018 ◽  
Author(s):  
Dale E. Matthews ◽  
Ralph S. Hill ◽  
Charles W. Bruny
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Fusion Energy ◽  
Life Cycle Management ◽  
Economic Benefits ◽  
Plant Design ◽  
Element Analysis ◽  
Energy Devices ◽  
The Future

ASME Nuclear Codes and Standards are used worldwide in the construction, inspection, and repair of commercial nuclear power plants. As the industry looks to the future of nuclear power and some of the new plant designs under development, there will be some significant departures from the current light water reactor (LWR) technology. Some examples are gas-cooled and liquid metal-cooled high temperature reactors (HTRs), small modular reactors (SMRs), and fusion energy devices that are currently under development. Many of these designs will have different safety challenges from the current LWR fleet. Variations of the current LWR technology are also expected to remain in use for the foreseeable future. Worldwide, many LWRs are planned or are already under construction. However, technology for construction of these plants has advanced considerably since most of the current construction codes were written. As a result, many modern design and fabrication methods available today, which provide both safety and economic benefits, cannot be fully utilized since they are not addressed by Code rules. For ASME Nuclear Codes and Standards to maintain and enhance their position as the worldwide leader in the nuclear power industry, they will need to be modernized to address these items. Accordingly, the ASME Nuclear Codes and Standards organizations have initiated the “2025 Nuclear Code” initiative. The purpose of this initiative is to modernize all aspects of ASME’s Nuclear Codes and Standards to adopt new technologies in plant design, construction, and life cycle management. Examples include modernized finite element analysis and fatigue rules, and incorporation of probabilistic and risk-informed methodology. This paper will present the vision for the 2025 ASME Nuclear Codes and Standards and will discuss some of the key elements that are being considered.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

Analytical Modelling of Elastic Interaction in Bolted Flange Gasketed Joints

2017 ◽  
Author(s):  
Linbo Zhu ◽  
Abdel-Hakim Bouzid ◽  
Jun Hong
Keyword(s):  
Analytical Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Experimental Tests ◽  
Elastic Interaction ◽  
Industrial Complex ◽  
Element Analysis ◽  
Minimum Number ◽  
Leakage Failure ◽  
Bolted Flange

Bolted flange joints are widely used in the fossil and nuclear power plants and other industrial complex. During their assembly, it is extremely difficult to achieve the target bolt preload and tightening uniformity due to elastic interaction. In addition to the severe service loadings the initial bolt load scatter increases the risk of leakage failure. The objective of this paper is to present an analytical model to predict the bolt tension change due to elastic interaction during the sequence of initial tightening. The proposed analytical model is based on the theory of circular beams on linear elastic foundation. The elastic compliances of the flanges, the bolts, and the gasket due to bending, twisting and axial compression are involved in the elastic interaction. The developed model can be used to optimize the initial bolt load tightening to obtain a uniform final preload under minimum number of tightening passes. The approach is validated using finite element analysis and experimental tests conducted on a NPS 4 class 900 weld neck bolted flange joint.

Improvement of FAC Maintenance Program Issued From BRT-CICERO™ via CFD Calculations

2016 ◽  
Author(s):  
Elodie Gipon
Keyword(s):  
Mass Transfer ◽  
Nuclear Power ◽  
Power Plants ◽  
Specific Effect ◽  
Piping System ◽  
Element Analysis ◽  
Accelerated Corrosion ◽  
Wear Rates ◽  
Maintenance Program ◽  
Calculation Algorithms

Flow Accelerated Corrosion (FAC) is very effective for nuclear power plant. This generalized corrosion can lead to the rupture of pipe and in some dramatic cases to casualties. During the last 20 years Electricité de France (EDF) has developed software called BRT-CICERO™ for the surveillance of the carbon steel piping system of its Nuclear Power Plants (NPPs). This software enables the operator to calculate the FAC wear rates by taking into account all the influencing parameters such as pipe isometrics, alloy content, chemical conditioning, design and operating parameters of the steam water circuit (temperature, pressure, etc…). This is a major tool to help operators organize their maintenance and inspections plan. The algorithms implemented in BRT-CICERO™ are based on tests conducted by EDF R&D, empirical results (national and international feedback), literature reviews and on permanent adjustments based on the operating feedback, via statistical studies. However, for some piping components, from the turbine’s hall, flow dynamics are not optimized and calculated FAC kinetics may be too conservative. EDF is committed for optimizing and increasing reliability of its maintenance programs to prevent the risk of pipe rupture due to FAC. As in consequence EDF is leading continuous improvement in parameters and calculation algorithms for BRT-CICERO™. Furthermore studies on the geometric characteristics of the pipes were conducted. In BRT-CICERO™ geometric effect of a pipe component (elbow reduction, tees …) is taken into account by considering a factor called “Geo” in the calculation to tune the thickness loss rate according the component type, its characteristics and specific effect on flow mass transfer. EDF implements finite element analysis software to compute the mass transfer coefficient k and so ascertain the “Geo” coefficient. These computed “Geo” coefficients are compared to those used in BRT-CICERO™. If necessary, current “Geo” coefficients used in BRT-CICERO™ will be adjusted and optimized to improve maintenance programs issued from the software. The presentation deals with the calculation method used for these studies and some results will be shown on tube and elbows.

Use of FEA Software in Evaluating Pump Vibration and Potential Remedial Actions to Abate Resonance in Industrial Vertical Pump/Motor Combinations

2016 ◽  
Author(s):  
Michael Tompkins ◽  
Robert Stakenborghs ◽  
Gregory Kramer
Keyword(s):  
Natural Frequency ◽  
Driving Force ◽  
Nuclear Power ◽  
Power Plants ◽  
Analysis Software ◽  
Element Analysis ◽  
Screw Pump ◽  
Twin Screw ◽  
Remedial Actions ◽  
Motor Structure

Reactor recirculation motor generator lube oil twin screw pumps are commonly found in nuclear power plants and throughout industry. In a vertical mounting configuration in which the electric motor is bolted atop the twin screw pump in an unsupported manner the natural frequency of the pump/motor structure can be quite low, resulting in damaging vibration. When a structure’s natural frequency coincides at or near the operating speed, or multiple thereof, a phenomena known as resonance can occur. Resonance can occur when a driving force, in this case minor imbalances in either the motor or pump, begins to vibrate and excite the structure resulting in greatly amplified levels of vibration. In this paper, finite element analysis software is utilized to first calculate the natural frequency of the pump/motor structure, and then potential modifications are modeled to determine their impact on eliminating harmful resonance.

Fatigue Effect of RCS Branch Line by Thermal Stratification

2003 ◽  
Author(s):  
Hag-Ki Youm ◽  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Tea-Eun Jin ◽  
Sun-Ki Lee ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Methodology ◽  
Element Analysis ◽  
Branch Line ◽  
Reactor Coolant ◽  
Stratification Effect ◽  
Fatigue Evaluation ◽  
Coolant System

Recent events reported at a number of nuclear power plants worldwide have shown that thermal stratification, cycling, and striping in piping can cause excessive thermal stress and fatigue on the piping material. These phenomena are diverse and complicated because of the wide variety of geometry and thermal hydraulic conditions encountered in reactor coolant system. Thermal stratification effect of re-branched lines is not yet considered in the fatigue evaluation. To evaluate the thermal load due to turbulent penetration, this paper presents a fatigue evaluation methodology for a branch line of reactor coolant system with the re-branch line. The locations of fatigue monitoring and supplemented inspections are discussed as a result of fatigue evaluations by Interim Fatigue Management Guideline (ITFMG) and detail finite element analysis. Although the revised CUF was increased less than 50 %, the CUF values for some locations was greater than the ASME Code limits.

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