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.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

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.

TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

2012 ◽  
Author(s):  
Roberta Ferri ◽  
Fulvio Mascari ◽  
Paride Meloni ◽  
Giuseppe Vella
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Heat Removal ◽  
Nuclear Fission ◽  
Dynamic Coupling ◽  
Hydraulic Conditions ◽  
Loss Of Coolant ◽  
Accident Condition

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants. The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA. The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents. Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal. The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes. The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.

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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