IDENTIFICATION OF IMPACT LOCATION IN A PLATE BASED ON ELASTODYNAMICS AND HIGHER ORDER TIME FREQUENCY METHOD

2008 ◽  
Vol 22 (09n11) ◽  
pp. 1331-1336
Author(s):  
S. K. LEE ◽  
S. J. KIM
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Group Velocity ◽  
Structural Damage ◽  
Nuclear Power ◽  
Thick Plate ◽  
Accurate Estimation ◽  
Frequency Method ◽  
Time Frequency ◽  
The Impact

In a nuclear power plant, impact force due to loose part is related to the structural damage in the plant. In general, the steam generator of the nuclear power plant is structured by thick plate. The paper presents a novel approach to locate an impact load in a thick plate. The approach is based on the analysis of the acoustic waveforms measured by a sensor array located on the plate surface. For accurate estimation of the location of the impact source, the time differences in the arrival times of the waves at the sensors and their propagation velocities are determined. The dispersion curves for multi modes of Lamb wave are calculated by using exact plate theory and SDPT. It is difficult to measure directly the group velocity for Lamb mode of acoustic waveform in the thick plate because they are dispersive wave. However, most of the energy in the wave is carried by the flexural waves (A0 mode), the group velocity of this mode is extracted using the CHOTF technique for estimating the impact source location. The estimates are shown to be in excellent agreement with the actual locations and it is applied to the damage analysis due to the loose part in a nuclear power plant.

IDENTIFICATION OF IMPACT LOCATION IN A NUCLEAR POWER PLANT BASED ON HIGHER ORDER TIME FREQUENCY ANALYSIS AND ELASTODYNAMICS

2008 ◽  
Vol 22 (11) ◽  
pp. 1025-1030 ◽  
Author(s):  
SANG KWON LEE ◽  
SU-GON KIM
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Group Velocity ◽  
Nuclear Power ◽  
Thick Plate ◽  
Plate Theory ◽  
Higher Order ◽  
Accurate Estimation ◽  
Time Frequency ◽  
The Impact

The paper presents a novel approach to locate an impact load in a thick plate. The approach is based on the analysis of the acoustic waveforms estimated at three different points on the plate surface. For accurate estimation of the location of the impact source, the time differences in the arrival times of the waves at the three points and their propagation velocities are determined. The dispersion curves for multi modes of Lamb wave are calculated by using exact plate theory. It is difficult to eatimate directly the group velocity for Lamb mode of acoustic waveform in the thick plate because they are dispersive wave. However, most of the energy in the wave is carried by the flexural waves (A0 mode), the group velocity of this mode is extracted using the CHOTF (combined higher order time frequency) technique for estimating the impact source location. The estimates are shown to be in excellent agreement with the actual locations and it is applied to the damage analysis due to the loose part in a nuclear power plant.

Locating Impact on Thick Plates by Acoustic Wave Analysis

2008 ◽  
Author(s):  
Ho-Wuk Kim ◽  
Sang-Kwon Lee
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Group Velocity ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thick Plate ◽  
Impact Load ◽  
Plate Theory ◽  
Thick Plates ◽  
The Impact

Loose parts in a steam generator of a nuclear power plant often impact the wall of the generator and become one of the damage sources in the nuclear power plant. In general, the steam generator of the nuclear power plant is structured by thick plates. This paper presents a novel approach to locating an impact load in a thick plate. The approach is based on an analysis of the acoustic waveforms measured by a sensor array located on the plate surface and theoretically obtained by either the exact elastodynamic or theory the approximate shear deformation plate theory (SDPT). For accurate estimation of the location of the impact source due to loose part, the time differences in the arrival times of the waves at the sensors and their propagation velocities are determined. This is accomplished through the use of a combined higher order time frequency (CHOTF) method, which is capable of detecting signals with lower signal to noise ratio compared to other available methods. The dispersion curves for multi modes of Lamb waves are calculated by using exact plate theory and SDPT. It is difficult to measure directly the group velocity for Lamb mode of acoustic waveform in the thick plate because they are dispersive waves. However, most of the energy in the wave is carried by the flexural waves (A0 mode); the group velocity of this mode is extracted by using the CHOTF technique for estimating the impact source location. The estimates are shown to be in excellent agreement with the actual locations and the technique is applied to the detection of the location of the impact load due to the loose part in a nuclear power plant.

scholarly journals Comparison and Evaluation of Domestic and Foreign Radiation Environmental Standards for Nuclear Power Plants

2021 ◽  
Vol 2083 (2) ◽  
pp. 022020
Author(s):  
Jiahuan Yu ◽  
Xiaofeng Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Discharge System ◽  
Radioactive Effluent ◽  
Effluent Discharge ◽  
The Impact

Abstract With the development of the nuclear energy industry and the increasing demand for environmental protection, the impact of nuclear power plant radiation on the environment has gradually entered the public view. This article combs the nuclear power plant radiation environmental management systems of several countries, takes the domestic and foreign management of radioactive effluent discharge from nuclear power plants as a starting point, analyses and compares the laws and standards related to radioactive effluents from nuclear power plants in France, the United States, China, and South Korea. In this paper, the management improvement of radioactive effluent discharge system of Chinese nuclear power plants has been discussed.

Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant

2006 ◽  
Author(s):  
Sang-Nyung Kim ◽  
Sang-Gyu Lim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Measurement Data ◽  
Significant Loss ◽  
Design Modification ◽  
Root Cause ◽  
Explosive Expansion ◽  
The Impact

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.

scholarly journals Blowout Accident Impact Analysis Method for the Siting of Offshore Floating Nuclear Power Plant in Offshore Oil Fields

2019 ◽  
Vol 2019 ◽  
pp. 1-7
Author(s):  
Zhigang Lan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Oil Spills ◽  
Oil Field ◽  
Oil Fields ◽  
Offshore Oil ◽  
The Impact

Focused on the utilization of nuclear energy in offshore oil fields, the correspondence between various hazards caused by blowout accidents (including associated, secondary, and derivative hazards) and the initiating events that may lead to accidents of offshore floating nuclear power plant (OFNPP) is established. The risk source, risk characteristics, risk evolution, and risk action mode of blowout accidents in offshore oil fields are summarized and analyzed. The impacts of blowout accident in offshore oil field on OFNPP are comprehensively analyzed, including injection combustion and spilled oil combustion induced by well blowout, drifting and explosion of deflagration vapor clouds formed by well blowouts, seawater pollution caused by blowout oil spills, the toxic gas cloud caused by well blowout, and the impact of mobile fire source formed by a burning oil spill on OFNPP at sea. The preliminary analysis methods and corresponding procedures are established for the impact of blowout accidents on offshore floating nuclear power plants in offshore oil fields, and a calculation example is given in order to further illustrate the methods.

On the Feasibility of Nuclear Versus Combustion Gas Turbine Propulsion for High-Speed Cargo Ships

2004 ◽  
Author(s):  
H. Boonstra ◽  
A. C. Groot ◽  
C. A. Prins
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Gas Turbine ◽  
High Speed ◽  
Nuclear Power ◽  
Parametric Design ◽  
Pebble Bed ◽  
Pressurized Water ◽  
Combustion Gas ◽  
The Impact

This paper presents the outcome of a study on the feasibility of a nuclear powered High-Speed Pentamaran, initiated by Nigel Gee and Associates and the Delft University of Technology. It explores the competitiveness of a nuclear power plant for the critical characteristics of a marine propulsion plant. Three nuclear reactor types are selected: the Pressurized Water Reactor (PWR), the Pebble-bed and Prismatic-block HTGR. Their characteristics are estimated for a power range from 100 MWth to 1000 MWth in a parametric design, providing a level base for comparison with conventional gas turbine technology. The reactor scaling is based on reference reactors with an emphasis on marine application. This implies that preference is given to passive safety and simplicity, as they are key-factors for a marine power plant. A case study for a 60-knot Pentamaran shows the impact of a nuclear power plant on a ship designed with combustion gas turbine propulsion. The Prismatic-block HTGR is chosen as most suitable because of its low weight compared to the PWR, in spite of the proven technology of a PWR. The Pebble-bed HTGR is considered too voluminous for High-Speed craft. Conservative data and priority to simple systems and high safety leads to an unfavorable high weight of the nuclear plant in competition with the original gas turbine driven Pentamaran. The nuclear powered ship has some clear advantages at high sailing ranges.

Numerical Simulation of Hydrogen Dispersion Inside a Compartment Using HYDRAGON Code

2016 ◽  
Author(s):  
M. Saeed ◽  
Yu Jiyang ◽  
B. X. Hou ◽  
Aniseh A. A. Abdalla ◽  
Zhang Chunhui
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Concentration Distribution ◽  
Nuclear Power ◽  
Turbulence Models ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Normal Operation ◽  
Accurate Estimation ◽  
Nuclear Power Plant Accident

During severe accident in the nuclear power plant, a considerable amount of hydrogen can be generated by an active reaction of the fuel-cladding with steam within the pressure vessel which may be released into the containment of nuclear power plant. Hydrogen combustion may occur where there is sufficient oxygen, and the hydrogen release rates exceed 10% of the containment. During hydrogen combustion, detonation force and short term pressure may be produced. The production of these gas species can be detrimental to the structural integrity of the safety systems of the reactor and the containment. In 1979, the Three Mile Island (1979) accident occurred. This accident compelled experts and researchers to focus on the study of distribution of hydrogen inside the containment of nuclear power plant. However after the Fukushima Dai-ichi nuclear power plant accident (2011), the modeling of the gas behavior became important topic for scientists. For the stable and normal operation of the containment, it is essential to understand the behavior of hydrogen inside the containment of nuclear power plant in order to mitigate the occurrence of these types of accidents in the future. For this purpose, it is important to identify how burnable hydrogen clouds are produced in the containment of nuclear power plant. The combustion of hydrogen may occur in different modes based on geometrical complexity and gas composition. Reliable turbulence models must be used in order to obtain an accurate estimation of the concentration distribution as a function of time and other physical phenomena of the gas mixture. In this study, a small scale hydrogen-dispersion case is selected as a benchmark to address turbulence models. The computations are performed using HYDRAGON code developed by Department of Engineering Physics, Tsinghua University, China. HYDRAGON code is a three dimensional thermal-hydraulics analysis code. The purpose of this code is to predict the behavior of hydrogen gas and multiple gas species inside the containment of nuclear power plant during severe accident. This code mainly adopts CFD models and structural correlations used for wall flow resistance instead of using boundary layer at a wall. HYDROGAN code analyzes many processes such as hydrogen diffusion condensation, combustion, gas stratification, evaporation, mixing process. The main purpose of this research is to study the influence of turbulence models to the concentration distribution and to demonstrate the code thermal-hydraulic simulation capability during nuclear power plant accident. The calculated results of various turbulence models have different prediction values in different compartments. The results of k–ε turbulence model are in reasonable agreement as compared to the benchmark experimental data.

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