A seismic analysis method for a block column gas-cooled reactor core

1979 ◽  
Vol 55 (3) ◽  
pp. 331-342 ◽  
Author(s):  
T. Ikushima ◽  
T. Nakazawa
Keyword(s):  
Seismic Analysis ◽  
Reactor Core ◽  
Analysis Method ◽  
Block Column

Core Seismic Experiment and Analysis of Hexagonal Bundle Model for Fast Reactor

2017 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Tomohiko Yamamoto ◽  
Seiji Kitamura ◽  
Shigeki Okamura
Keyword(s):  
Seismic Response ◽  
Fast Reactor ◽  
Impact Force ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Three Dimensions ◽  
Analysis Method ◽  
Core Element ◽  
Core Elements ◽  
Advanced Analysis

To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Therefore, advanced analysis method must be developed to calculate seismic response of a fast reactor core. The fast reactor core is generally made of several hundred core elements which are hexagonal flexible beams embedded at the lower support plate in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement, vertical displacement (rising) and impact force of each core element may cause a trouble for control rod insertability and core element intensity. Therefore, a seismic analysis method of a fast reactor core considering three-dimensional nonlinear behavior, such as impact, fluid-structure interaction, was developed. 1/1.5 scale 37 core element mock-ups hexagonal-matrix experiment was performed to validate the core elements vibration analysis code in three dimensions (REVIAN-3D). Vertical behavior (rising displacement) and horizontal behavior (impact force) were good agreement with experiments by the validation of REVIAN-3D.

Fast Reactor Core Seismic Analysis for Verification of Assessment Model of Control Rod

2018 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Kazuteru Kawamura ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fuel Assembly ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Assessment Model ◽  
Guide Tube ◽  
Analysis Method ◽  
Control Rod ◽  
Support Plate ◽  
The Core ◽  
Core Elements

The control rod guide tube self-stands on the core support plate. The control rod is inserted in the control rod guide tube, and the control rod hangs from the upper structure of the reactor. At scrum in case of an earthquake, the control rod is detached and it sits on the seating structure in the control rod guide tube (Fig.1). In a vertical earthquake, the control rod guide tube is raised from the core support plate, and the control rod is also raised from the control rod guide tube. Therefore, drawing out may arise. During the earthquake after scrum, the rising behavior is different from the other core elements because the control rod and the control rod guide tube rise interfering each other. The control rod guide tube is raised more easily than the fuel assembly by the vertical differential pressure of the core during operation, because the control rod guide tube is lighter than the fuel assembly. Therefore, it is necessary to restrain the rising of the control rod guide tube. The sleeve dashpot structure, in which a sleeve is attached on the upper surface of the receptacle tube, is employed. Moreover, the control rod guide tube is equipped with the control rod dashpot in order to restrain the rising displacement of the control rod. This paper summarizes the analysis method of the rising behavior of the single control rod guide tube and the rising behavior of the control rod and the control guide tube after the control rod is inserted.

scholarly journals Development of seismic analysis method of fast reactor core

2016 ◽  
Vol 82 (839) ◽  
pp. 16-00093-16-00093
Author(s):  
Akihisa IWASAKI ◽  
Kazuo HIROTA ◽  
Masatsugu MONDE ◽  
Iwao IKARIMOTO
Keyword(s):  
Fast Reactor ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Analysis Method

Development of a Core Seismic Analysis Method for a Fast Reactor

2016 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Kazuo Hirota ◽  
Masatsugu Monde ◽  
Shinichiro Matsubara ◽  
Iwao Ikarimoto
Keyword(s):  
Fast Reactor ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Great Earthquake ◽  
Analysis Method ◽  
Core Element ◽  
The Core ◽  
Vibration Behavior ◽  
Core Elements ◽  
Core Seismic Analysis

A fast reactor core consists of several hundreds of core assemblies, which are hexagonal flexible beams embedded at the lower support plate in a hexagonal arrangement, separated by small gaps, and immersed in a fluid. Core assemblies have no support for vertical fixing in order to avoid the influence of thermal expansion and swelling. These days, in Japan, it has become necessary to postulate huge earthquakes in seismic evaluations. If a great earthquake occurs, the large displacement and impact force in each core assembly may cause problems with control rod insertability and core assembly strength. So, it is necessary to grasp the vibration behavior of the core elements during an earthquake in order to appropriately design the core support structures and core elements of a fast reactor. Thus, considering horizontal and vertical forces (impact forces and fluid forces) acting on the core elements during an earthquake, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacements (rising). This paper summarizes the details of the core element vibration analysis code in 3D (REVIAN-3D) that has been developed.

scholarly journals Seismic Analysis Method for Cabinets and Other Components of a Process Computer System.

1997 ◽  
Vol 63 (614) ◽  
pp. 3361-3366
Author(s):  
Hiroshi NIWA ◽  
Hidehiro FUKUI ◽  
Youichi SASAKI ◽  
Fumio HARA
Keyword(s):  
Computer System ◽  
Seismic Analysis ◽  
Analysis Method ◽  
Process Computer

A Safety Evaluation of HTTR Core Components Against 2011 Tohoku Earthquake

2013 ◽  
Author(s):  
Kazuhiko Iigaki ◽  
Masato Ono ◽  
Yosuke Shimazaki ◽  
Daisuke Tochio ◽  
Atsushi Shimizu ◽  
...  
Keyword(s):  
Structural Integrity ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Safety Evaluation ◽  
Tohoku Earthquake ◽  
Reactor Power ◽  
2011 Tohoku Earthquake ◽  
The 2011 Tohoku Earthquake ◽  
Maximum Acceleration ◽  
Core Components

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.

Research on the Seismic Analysis and Optimization of High Flux Reactor Nuclear Power Piping System Based on Multi-Point Response Spectrum Method

2020 ◽  
Author(s):  
Xuan Huang ◽  
Pingchuan Shen ◽  
Shuai Liu ◽  
Jian Liu ◽  
Xiaozhou Jiang ◽  
...  
Keyword(s):  
Seismic Analysis ◽  
Response Spectrum ◽  
High Flux ◽  
Analysis Method ◽  
Response Spectrum Method ◽  
High Flux Reactor ◽  
Reactor Coolant ◽  
Spectrum Method ◽  
Flux Reactor ◽  
Coolant System

Abstract High flux reactor is an important engineering test reactor, which can be used in irradiation research of materials, chemistry, isotopes, medicine and other fields. In the high flux reactor coolant system, there are a large number of nuclear pipes and the layout is complex. The optimization of seismic analysis method for reactor coolant system is an important part in the design process to ensure the nuclear pipes meet the design specifications. The traditional single point response spectrum method needs to envelope the response spectrum of different floors as the analysis input. This method is difficult to give the reasonable seismic load to the numerous nuclear pipes and it will increase the design cost and the difficulty of safety analysis about nuclear pipe. In this paper, an optimized seismic analysis method of reactor coolant system is proposed. By using the multi-point response spectrum method, the optimization of different excitation loading modes for different constrained support points is realized. The analysis results show that the multi-point response spectrum method can solve the problem that different support points are located at different elevation floors in the reactor coolant system, which makes the calculation results more accurate and reasonable. Compared with the traditional method, it can make the design more efficient and practical.

A Large-Scale Numerical Simulation on Bubbly and Liquid Film Flows in Narrow Fuel Channels

2005 ◽  
Author(s):  
K. Takase ◽  
H. Yoshida ◽  
Y. Ose ◽  
H. Akimoto
Keyword(s):  
Large Scale ◽  
Two Phase Flow ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Phase Flow ◽  
Analysis Method ◽  
Two Phase ◽  
Light Water ◽  
Water Reactor ◽  
Fuel Bundle

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were carried out using a newly developed two-phase flow analysis method and a highly parallel-vector supercomputer. Conventional analysis methods such as subchannel codes need composition equations based on many experimental data. Therefore, it is difficult to obtain highly prediction accuracy on the thermal design of the advanced light-water reactor core if the experimental data are insufficient. Then, a new analysis method using the large-scale direct numerical simulation of water-vapor two-phase flow was proposed. The coalescence and fragmentation of small bubbles were investigated numerically and the bubbly flow dynamics in narrow fuel channels were clarified. Moreover, the liquid film flow inside a tight-lattice fuel bundle which is used to the advanced light-water reactor core was analyzed and the water and vapor distributions around fuel rods and a spacer were estimated quantitatively.

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