Seismic Test Analysis Evaluation of Motor-Operated Valve Actuators for Nuclear Power Plant

2018 ◽  
Author(s):  
Nobuo Kojima ◽  
Yoshitaka Tsutsumi ◽  
Kazuyoshi Yonekura ◽  
Koji Nishino ◽  
Yukio Watanabe ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Ground Motions ◽  
Seismic Resistance ◽  
Test Results ◽  
Analysis Method ◽  
The Finite Element Method ◽  
Test Analysis ◽  
High Acceleration ◽  
Compression Characteristics

The function of motor-operated valves provided in nuclear power plants during an earthquake has been previously evaluated on the basis of seismic test results; however, since the response acceleration has increased with a recent reassessment of standard earthquake ground motions, it is necessary to evaluate their function during high acceleration earthquakes. In order to carry out seismic test analysis, we examined the compression characteristics of the gaskets in three kinds of representative valves, in terms of the current load and slide coefficient, by element tests. We created an analysis method using the finite element method which used the gasket’s characteristics, and the result of these analyses could explain the loosing and the amplitude of the bolt connecting force and gasket slipping at the time of excitation to the safety side. According to the analysis, the actuator connection part of the motor-operated butterfly valve showed a seismic resistance of 20 × 9.8m/s2 and other parts showed a seismic resistance of 30 × 9.8m/s2.

Selection of the Test Specimens for Seismic Tests of Air-Operated Valve Actuators for Nuclear Power Plants

2019 ◽  
Author(s):  
Nobuo Kojima ◽  
Yoshitaka Tsutsumi ◽  
Yoshinao Matsubara ◽  
Koji Nishino ◽  
Yasuyuki Ito ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ground Motions ◽  
Test Results ◽  
Earthquake Ground Motions ◽  
Dynamic Function ◽  
Design Earthquake ◽  
Seismic Tests ◽  
Selection Of

Abstract The soundness for the function of air-operated valves in nuclear power plants during earthquake has been investigated via seismic test results and so forth. Since the seismic response acceleration has increased more and more with a recent reassessment of design earthquake ground motions conducted according to the revised Japanese nuclear safety regulation, it is necessary to evaluate the soundness for the function of various valves subject to large acceleration beyond design basis. The air-operated valves currently installed in the nuclear power plants in Japan play the important roles in the sever accident events. In this study, we classified them based on the valve type, manufactures and the previous test results. Furthermore, we proposed the strategy for evaluating the seismic-proof and the seismic test condition for examining the soundness of the dynamic function. Here, the dynamic function is defined as the function required under and after earthquakes.

Calculation of Seismic Resistance of Group «B» Cranes Operating at Nuclear Power Plants (NPP)

2018 ◽  
Vol 12 (3) ◽  
pp. 73-80
Author(s):  
Y.I. Pimchin ◽  
V.A. Naugolnov ◽  
G.A. Naumenko ◽  
I.U. Pimchin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Seismic Resistance ◽  
Group B

Fiber-Optic Transmitter for Nuclear Power Plants

2004 ◽  
Author(s):  
Junichi Higashi ◽  
Shinichi Murakawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fiber Optic ◽  
Optic Fiber ◽  
Test Results ◽  
Maintenance System ◽  
Fiber Technology ◽  
Fabry Perot ◽  
Validation Tests

A promising Fiber-Optic Differential Pressure (DP) Transmitter is under development in Flexible Maintenance System (FMS) Projects that supported by Ministry of Economic, Trade, and Industries of Japan. The object of FMS projects is to improve maintenance works at nuclear power plants with latest technology. The new DP Transmitter uses optic-fiber technology of Extrinsic Fabry-Perot Sensor and Fizeau White-Light Cross-Correlator. Validation tests were performed to evaluate the tolerance of the DP transmitter in Nuclear Power Plant conditions. General requirements of PWR are accuracy (repeatability and linearity) of within +/−0.5%, pressure-proof of maximum 17.16MPa, Irradiation of 100Gy, and temperature range of 10–50 degrees centigrade at normal condition. The test results show the new DP transmitter can be expected as the next generation instrumentation in Nuclear Power Plants.

Development of Advanced Fire Zone Model Applicable to Fire PRA for Nuclear Power Plant

2020 ◽  
Author(s):  
Junghoon Ji ◽  
Koji Shirai ◽  
Koji Tasaka ◽  
Toshiko Udagawa
Keyword(s):  
Oxygen Concentration ◽  
Nuclear Power ◽  
Power Plants ◽  
Fire Risk ◽  
Zone Model ◽  
Radiant Heat Transfer ◽  
Test Results ◽  
Compartment Fire ◽  
Jet Flame ◽  
Fire Source

Abstract In implementing the fire PRA for nuclear power plants, a highly predictive fire model is required for more realistic fire scenarios and fire risk assessment. The fire simulation zone model BRI2002 developed in Japan has been continuously improved to allow analysis considering the characteristics of a compartment fire. In this study, a heat feedback phenomenon was introduced in BRI2002, in which combustion of a fire source can be accelerated by radiant heat transfer inside the compartment during a compartment fire. Not only the thermal radiation from the flame and smoke layers, but also radiation from the hot ceiling surface and the ceiling jet flame were considered when the flame impinges with the ceiling. In addition, in the zone model, the existing model for predicting the oxygen concentration in a compartment was improved so that the oxygen concentration could be predicted considering the vertical location of a fire source (height from the floor). The prediction results were verified by full-scale compartment fire test results. As a result of the calculation in which the fire source is installed at 2 m above the floor, the prediction results for the burning rate and zone temperature were well consistent with the test results.

Seismic Design of Free Standing Racks in Japanese Nuclear Power Plants

2017 ◽  
Author(s):  
Yu Takaki ◽  
Katsuhiko Taniguchi ◽  
Junichi Kishimoto ◽  
Akihisa Iwasaki ◽  
Yoshitsugu Nekomoto ◽  
...  
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Method ◽  
Seismic Resistance ◽  
Spent Fuel ◽  
Analysis Model ◽  
Free Standing ◽  
Scale Free ◽  
Seismic Design Method

The free standing racks are spent fuel storage racks with self-sustained structure without fixation to the pit floor or pit walls. If a free standing rack receives a force to move it due to an earthquake, the force acting on each member of the rack is reduced in compared to the floor-anchored racks owing to sliding of the free standing rack. Now it is planned to exchange the existing floor-anchored racks with the free standing racks to secure higher seismic resistance. In previous studies, efforts were made to establish a behavior analysis model that allows for evaluation of sliding and rocking behaviors of free standing racks and to make out a seismic design method based on an evaluation technique to evaluate, in a conservative manner, vibration test results of full-scale free standing racks. The free standing racks which consist of connected eight racks are designed with this seismic design method. It was confirmed that the free standing racks have enough seismic resistance by performing evaluation using the basic seismic motion and making an analysis on beyond the design event.

Fluid-Structure Dynamics With a Modal Hybrid Method

1992 ◽  
Vol 114 (2) ◽  
pp. 133-138 ◽  
Author(s):  
B. Brenneman ◽  
M. K. Au-Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Hybrid Approach ◽  
Dynamic Responses ◽  
Test Results ◽  
Strongly Coupled ◽  
Large Structures ◽  
Modal Synthesis ◽  
Loss Of Coolant ◽  
Thermal Shields

Large structures in nuclear power plants are often separated by very thin fluid-filled cavities. For example, core support structures, thermal shields, and reactor vessels are usually large concentric cylindrical shells with annuli between them as small as 2 percent of the shell diameter. Such thin cavities cause the structures to be very strongly coupled, and such coupling must be accurately modeled to predict the dynamic responses of new designs to turbulence, pump acoustic loading, loss-of-coolant accidents, and seismic events. This paper summarizes a very versatile and efficient method of solving these problems with small personal computers. Among other things, this method uses component modal synthesis with the hybrid approach, and the solution of the resulting unsymmetric eigenvalue problem for the coupled vibration modes. System responses are then found in terms of “right” and “left” eigenvectors. Comparisons with test results are also presented.

Research of Ageing and Life Assessment Methods of Power Transformers in Nuclear Power Plants

2012 ◽  
Vol 614-615 ◽  
pp. 1109-1115 ◽  
Author(s):  
Cheng Feng Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Assessment Method ◽  
Thermal Ageing ◽  
Degree Of Polymerization ◽  
Transformer Oil ◽  
Life Assessment ◽  
Power Transformers ◽  
Analysis Method

The effective methods of the ageing and life assessment for large and medium-sized power transformers used in nuclear power plants are analyzed and described, including the thermal ageing life assessment method for transformer solid insulation, the gas analysis method of CO and CO2 in the transformer oil, the average degree of polymerization method, furfural content analysis method, and the analysis method based on the insulation ageing-related electrical parameters. The analysis results show that the methods used can reasonably assess the remaining life of the transformers. These methods have important reference value to the ageing and life management for the large and medium-sized power transformers in nuclear power plants.

Pressure Tests for Thickness Management of Wall Thinning Tees

2012 ◽  
Author(s):  
Kaina Teshima ◽  
Yoichi Iwamoto ◽  
Kiminobu Hojo ◽  
Tomoyuki Oka ◽  
Kunihiro Kobayashi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Outer Diameter ◽  
Test Results ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Pressure Tests ◽  
Design Pressure

Although the minimum thickness of pipe wall required (tsr) of T-joints (tees) of class 2, 3 and lower classes of nuclear power plants in Japan is calculated from the design pressure and temperature, there is no rule or standard of wall thinning T-joints for thickness management. This paper describes the pressure tests procedure and six test results with parameters of T-joint geometry such as outer diameter D, thickness T and T/D to establish structural integrity of wall thinning T-joints. Based on the fracture surface observation, a ductile crack initiation of each test mock-ups was confirmed.

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