Dynamic Response of a Nuclear Power Plant Subjected to External Accident Event

2010 ◽  
Author(s):  
Rosa Lo Frano ◽  
Giuseppe Forasassi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Containment System ◽  
Dynamic Loadings ◽  
Water Reactors ◽  
Set Up

In recent times there is a renewed worldwide interest in the development and application of advanced nuclear power plants (NPPs). Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR in Finland and France, AP1000 in China, etc.) and more are under consideration for licensing in several countries. Innovative NPPs are designed to be built with very broad siting conditions; therefore the safety aspects related to the external events might follow new scenarios and failure modes, different from those well known for the currently operated reactors. In this paper, the intent is evaluating the structural integrity of a nuclear containment system subjected to dynamic loadings due to a Design Base Earthquake and an aircraft impact (large size civilian jets or military aircrafts impact), which represent the two most relevant external accidents that should be considered and investigated as part of the basic design of a NPP in particular a III+ and IV Gens. In fact a suitable safety design of the NPP containment system (according to the international safety and design code guidelines, as NRC or IAEA ones), even if designed to meet other design goal, may represent a “built-in protection” to avoid or mitigate the effects of mentioned dynamic loadings. To the purpose a rather sophisticated numerical methodology, adopting finite element (FEM) approach, is employed for studying the overall dynamic behaviour of nuclear reactor and to determine the structural effects of the propagation of dynamic seismic as well as impulsive loads (containment structure response) up to the relevant nuclear components. Therefore representative three-dimensional FEM models of mentioned NPP containment and aircraft structures were set up, and used, in the performed analyses taking also into account the suitable materials behaviour and their related constitutive laws as well as the seismic excitation (determined according to the NRC rules). Moreover the performed analyses and the carried out response analyses of internal components, to both the ground motion and impact loads, were studied to check the considered NPP containment strength reserve in the case of the considered events. The obtained results seem to confirm the possibility to achieve an optimization of the NPP internal components.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

An Assessment of Mechanical Properties of A508-3 Steel Used in Chinese Nuclear Reactor Vessels

2014 ◽  
Author(s):  
Meifang Yu ◽  
Zhen Luo ◽  
Y. J. Chao
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Construction Materials ◽  
Under Construction ◽  
Made In China ◽  
Phase Out

China has very ambitious goals of expanding its commercial nuclear power by 30 Giga-Watts within the decade and wishes to phase out fossil fuels emissions by 40–45% by 2020 (from 2005 levels). With over 50 new nuclear power plants under construction or planned and a design life of 60 years, any discussions on structural integrity become very timely. Although China adopted its nuclear technology from France or US at present time, e.g. AP1000 of Westinghouse, the construction materials are primarily “Made in China”. Among all issues, both the accumulation of the knowledge base of the materials and structures used for the power plant and the technical capability of engineering personnel are imminent. This paper attempts to compile and assess the mechanical properties, Charpy V-notch impact energy, and fracture toughness of A508-3 steel used in Chinese nuclear reactor vessels. All data are collected from open literature and by no means complete. However, it provides a glimpse into how this domestically produced steel compares with western reactor vessel steels such as US A533B and Euro 20MnMoNi55.

Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension

1989 ◽  
Vol 111 (3) ◽  
pp. 234-240 ◽  
Author(s):  
G. Yagawa ◽  
Y. Ando ◽  
K. Ishihara ◽  
T. Iwadate ◽  
Y. Tanaka
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Behavior ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Test Results ◽  
Unstable Crack ◽  
Pressurized Thermal Shock ◽  
Series Of Experiments

An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ integrals.

scholarly journals Anthropogenic Radiocarbon: Past, Present, and Future

Radiocarbon ◽  
1986 ◽  
Vol 28 (2A) ◽  
pp. 668-672 ◽  
Author(s):  
Pavel Povinec ◽  
Martin Chudý ◽  
Alexander Šivo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Pressurized Water Reactors ◽  
Anthropogenic Radionuclides ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Weapon Testing ◽  
Water Reactors

14C is one of the most important anthropogenic radionuclides released to the environment by human activities. Weapon testing raised the 14C concentration in the atmosphere and biosphere to +100% above the natural level. This excess of atmospheric C at present decreases with a half-life of ca 7 years. Recently, a new source of artificially produced 14C in nuclear reactors has become important. Since 1967, the Bratislava 14C laboratory has been measuring 14C in atmospheric 14CO2 and in a variety of biospheric samples in densely populated areas and in areas close to nuclear power plants. We have been able to identify a heavy-water reactor and the pressurized water reactors as sources of anthropogenic 14C. 14C concentrations show typical seasonal variations. These data are supported by measurements of 3H and 85Kr in the same locations. Results of calculations of future levels of anthropogenic 14C in the environment due to increasing nuclear reactor installations are presented.

FE Simulation of Welding Residual Stresses in Reactor Vessel Head Penetrations

2009 ◽  
Author(s):  
Anne Teughels ◽  
Christian Malekian
Keyword(s):  
Residual Stresses ◽  
Nuclear Power ◽  
Power Plants ◽  
Element Model ◽  
Mechanical Analysis ◽  
Alloy 600 ◽  
Finite Elements Analysis ◽  
Pressurized Water ◽  
Water Reactors ◽  
Set Up

The penetrations in the early Pressurized Water Reactors Vessels are characterized by Alloy 600 tubes, welded by Alloy 182/82. The Alloy 600 tubes have been shown to be susceptible to PWSCC (Primary Water Stress Corrosion Cracking) which may lead to crack forming. The cracking mechanism is driven mainly by the welding residual stresses and, in a second place, by the operational stresses in the weld region. It is therefore of big interest to quantify the weld residual stresses correctly. In order to determine the welding residual stresses, the weld procedure is simulated numerically by finite elements analysis. In the article, central as well as eccentric sidehill nozzles on the vessel head are analyzed. For the former a 2-dimensional axisymmetrical finite element model is used, whereas for the latter a 3-dimensional model is set up. A nonlinear transient thermo-mechanical analysis is performed, which is preceded by a transient thermal analysis simulating the heating during the multipass welding. Weld beads are deposited “all-at-once”. Different positions on the vessel head are compared and the influence of the sidehill effect is illustrated. The methodology is applied to the reactor vessels of the Belgian nuclear power plants by Tractebel Engineering (Belgium). The results are compared with literature. The global approach in both cases is very similar but is applied to different configurations, specific for each plant.

Validation of a CFD Code for the Analysis of Hydrogen Behaviors and Thermal Hydraulics in Containments

2018 ◽  
Author(s):  
Meilan Chen ◽  
Zeming Zheng
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Convection Heat Transfer ◽  
Navier Stokes ◽  
Thermal Hydraulics ◽  
Navier Stokes Equations ◽  
Water Metal ◽  
Water Reactors

During the process of core melt-down accident in light water reactors, large quantities of hydrogen generated by drastic water-metal reaction are released to the containment. Subsequently, hydrogen-rich layer may be formed under the dome of the containment, threatening the integrity of nuclear Power Plants (NPPs). In the framework of a China national R&D project, China Nuclear Power Research Institute (CNPRI) has developed a three dimensional CFD Code for the assessment of hydrogen behaviors and relative thermal hydraulics in containment. The code solves the time-dependent Navier-Stokes Equations with multi-gas species. Validation with International Standard Problems (ISP) and other test data based on a Phenomena Identification and Ranking Table (PIRT) has been undergoing together with the development of this code. In this paper, the test cases of HYJET, COPAIN and TOSQAN 101 Test are validated. Stratification, buoyancy induced mixing in gases, convection heat transfer and condensation on surface are evaluated in the former two cases, while gas entrainment and mixing by spray droplets in the later one. Excellent agreements between experimental data and model predictions are obtained. In order to meet the requirements for application of the code in practical NPP design and safety analysis, further validations of other phenomena in PIRT should be performed in the near future.

Nuclear Power Plant Fires and Explosions: Part I — Plant Designs and Hydrogen Ignition

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Reactor Power ◽  
Pressurized Water ◽  
Hydrogen Ignition ◽  
History Of ◽  
Water Reactors

Requiring further investigation, hydrogen explosions and fires have occurred in several operating nuclear reactor power plants. Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. This paper is the first paper in a series of publications to discuss this issue. In particular, the different types of reactors that have a history of fires and explosions are discussed here, along with a discussion of hydrogen generation in commercial reactors, which provides the fuel for fires and explosions in nuclear power plants. Overall, this paper is a review of pertinent information on reactor designs that is of particular importance to this multi-part discussion of hydrogen fires and explosions. Without a review of reactor designs and hydrogen generation, the ensuing technical discussions are inadequately backgrounded. Consequently, the basic designs of pressurized water reactors (PWR’s), boiling water reactors (BWR’s), and pressure-tube graphite reactors (RBMK) are discussed in adequate detail. Of particular interest, the Three Mile Island design for a PWR is presented in some detail.

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