Seismic Safety Margin of Cylindrical Liquid Storage Tanks in Nuclear Power Plants

2011 ◽  
Author(s):  
Akira Maekawa ◽  
Tsuneo Takahashi ◽  
Katsuhisa Fujita
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Scale ◽  
Safety Margin ◽  
Buckling Load ◽  
Storage Tanks ◽  
Seismic Safety ◽  
Liquid Storage ◽  
Design Value

In Japanese nuclear power plants, quantitative evaluation for seismic safety margin of the equipment is an important issue. In this study, the seismic safety margin of cylindrical liquid storage tanks used in nuclear power plants was investigated experimentally and analytically using test tanks with water inside. The buckling load of the tanks was examined because buckling was their dominant damage mode. The test tanks were reduced-scale models similar to the large-scale liquid storage tanks used in nuclear power plants. The experimental buckling load was compared with the design value. Furthermore, dynamic and static elastic-plastic buckling simulations by finite element analysis using a three-dimensional model were made and then the simulation results were compared with the experimental and design values. The simulated and experimental results agreed well, showing the values were the nearly-true buckling load, that is, proof stress. The design value was lower than the other values, indicating the difference was the seismic safety margin. The above results illustrated that existing actual tanks would have a bigger seismic safety margin.

Investigation on Buckling Behavior of Cylindrical Liquid Storage Tanks Under Seismic Excitation: 1st Report — Investigation on Elephant Foot Bulge

2003 ◽  
Author(s):  
Tomohiro Ito ◽  
Hideyuki Morita ◽  
Koji Hamada ◽  
Akihisa Sugiyama ◽  
Yoji Kawamoto ◽  
...  
Keyword(s):  
Storage Tank ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Storage Tanks ◽  
Scaled Model ◽  
Liquid Storage ◽  
Shear Buckling ◽  
Thin Walled ◽  
Liquid Storage Tank

When a thin walled cylindrical liquid storage tank suffers a large seismic base excitation, buckling phenomena may be caused such as bending buckling at the bottom portion and shear buckling at the middle portion of the tank. However, the dynamic behaviors of the tanks is not fully clarified, especially those from the occurrence of buckling to some failures. In this study, bending buckling phenomena were focused which will be categorized as diamond buckling and elephant foot bulge. As ones of a series of studies, dynamic buckling tests were performed using large scale liquid storage tank models simulating thin walled cylindrical liquid storage tanks in nuclear power plants. The input seismic acceleration was increased until the elephant foot bulge occurred, and the vibrational behavior before and after buckling was investigated. In addition to the large scaled model tests, fundamental tests using small scaled tank models were also performed in order to clarify the effects of dynamic liquid pressure on the buckling threshold and deformation patterns.

Failure Probability of Degraded Pipes Based on Probabilistic Fracture Mechanics for Seismic Safety Margin Assessment on NPP

2010 ◽  
Author(s):  
Yinsheng Li ◽  
Masaki Nakagawa ◽  
Katsumi Ebisawa ◽  
Shinobu Yoshimura ◽  
Hiroyuki Kameda
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ground Motions ◽  
Safety Margin ◽  
Mechanical Analysis ◽  
Margin Assessment ◽  
Seismic Safety ◽  
Design Values

The Niigata-ken Chuetsu-oki Earthquake happened in July 2007. Although the observed seismic ground motions of Kashiwazaki-Kariwa nuclear power plants greatly exceeded their design values, the important functions of the plants such as “shutdown”, “cooling”, and “confinement” were successfully ensured. Therefore, assessment of the seismic safety margin of nuclear power plants, both their systems and their components, becomes a very important issue. In this paper, failure probability of cracked pipes in existing nuclear power plants is analyzed by employing an improved probabilistic fracture mechanical analysis code and considering stress corrosion cracking and fatigue. Based on the analysis results of the failure probability and the definitions of the seismic safety margin, the seismic safety margin of degraded pipes in existing nuclear power plants is investigated.

Planning of Large-Scale In-Situ Gas Generation Experiment in Korean Radioactive Waste Repository

2010 ◽  
Author(s):  
Juyoul Kim ◽  
Sukhoon Kim ◽  
Jin Beak Park ◽  
Sunjoung Lee
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Scale ◽  
Radioactive Waste Disposal ◽  
Gas Generation ◽  
Radioactive Waste Repository ◽  
Waste Repository

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at Gyeongju city, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that affect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility. Gas is generated in association with these processes and has the potential threat to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction-oxidation reactions. The Korean research project on gas generation is being performed by Korea Radioactive Waste Management Corporation (hereafter referred to as “KRMC”). A full-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be done as an in-depth study during ten years at least. In order to examine gas generation issues from an LILW repository which is being constructed and will be completed by the end of December, 2012, two large-scale facilities for the gas generation experiment will be established, each equipped with a concrete container carrying on 16 drums of 200 L and 9 drums of 320 L of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be fully filled with ground water that provides representative geochemical conditions and microbial inoculation in the near field of repository. In the experiment, the design includes long-term monitoring and analyses for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual periodical sampling. A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.

The Position Design for M310 Reactor Vessel’s Instrumentation Penetration Based on None-Alignment Analysis

2013 ◽  
Author(s):  
Xin Xia ◽  
Hua Du ◽  
Ning Li ◽  
Bin Xu ◽  
Yan Li ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Economic Losses ◽  
Design Value ◽  
Alignment Analysis ◽  
Reactor Pressure ◽  
Reactor Internals

The position of the reactor pressure vessel’s instrumentation penetration is the most important factor that influences the none-alignment between the instrumentation penetration and the instrumentation tube of reactor internals. The more the none-alignment, the more wear the flux thimble will suffer, which will damage the flux detector and cause economic losses. This paper analyzes the none-alignment between the instrumentation penetration and the instrumentation tube, and combines the feedback of the experience in nuclear power plants having been built or being build, puts forward the reasonable design value for the instrumentation penetration’s position.

Study on Dynamic Strength Evaluation Method of Mechanical Members Based on Energy Balance

2007 ◽  
Author(s):  
Keisuke Minagawa ◽  
Satoshi Fujita ◽  
Seiji Kitamura ◽  
Shigeki Okamura
Keyword(s):  
Energy Balance ◽  
Experimental Model ◽  
Ultimate Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Dynamic Strength ◽  
Input Energy ◽  
Seismic Safety ◽  
Strength Evaluation

This paper describes the dynamic strength evaluation of piping installed in nuclear power plants from a viewpoint of energy balance. Mechanical structures installed in nuclear power plants such as piping and equipment are usually designed statically in elastic region. Although these mechanical structures have sufficient seismic safety margin, comprehending the ultimate strength is very important in order to improve the seismic safety reliability in unexpected severe earthquakes. In this study, ultimate strength of a simple single-degree-of-freedom model is investigated from a viewpoint of energy balance equation that is one of valid methods for structural calculation. The investigation is implemented by forced vibration experiment. In the experiment, colored random wave having predominant frequency that is similar to natural frequency of the experimental model is input. Stainless steel and carbon steel are selected as material of experimental model. Excitation is continued until the experimental model is damaged, and is carried out with various input levels. As a result of the experiment, it is confirmed that input energy for failure increase with an increase of time for failure. Additionally it is confirmed that input energy for failure depend on the material.

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