• Journal of the Korean Society of Manufacturing Process Engineers
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• v.18 no.5
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• pp.53-59
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• 2019

Continuous ship unloaders (CSUs) are used for the uninterrupted transport of material in processing industries, power plants, and harbors in accordance with the stream rate of the material. This study analyzed the structural integrity and fatigue life of a CSU structure using finite element structural analysis in ANSYS APDL software. The stress varied greatly depending on the luffing angle and the slew angle of the boom conveyor. The structural integrity of the CSU girder was evaluated by applying ASME BPVC Section VIII Division 2. The fatigue cycle at the angle with the greatest stress difference was calculated. The fatigue cycle was calculated by applying the JIS B 8821:2013 fatigue curve. It was confirmed that the fatigue cycle of the CSU satisfies the allowable fatigue of 200,000 cycles.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.555-561
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• 1994

The program of performing a fast and automatic analysis of gamma ray spectra obtained by a Multi-Channel Analyzer (MCA) is developed for the surveillance of the nuclear fuel integrity. The integrity of the nuclear fuel is confirmed by the measurement of the radiation level of the reactor coolant through the real time monitoring and the periodic sampling analysis. In Yonggwang nuclear power plane 3 and 4, the Process Radiation Monitoring System (PRMS), which is a real time monitoring system, provides a measure of the fuel integrity. Currently, its spectrometer channel can identify only one radionuclide at a time since the signal processing unit of the spectrometer channel is a Single Channel Analyzer (SCA). To improve the PRMS, it is necessary to substitute the MCA for the SCA The program is operated in a real time mode and an on-demand mode, and automatically performed for all procedures. The test results by using the National Bureau of Standards (NBS) mixed standard source are in good agreement with those from Canberra System 100 which is a commercial MCA Consequently, the developed program seems to be employed for automatic monitoring of gamma rays in nuclear power plants.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.636-642
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.268-268
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• 1999

• Journal of the Korea Academia-Industrial cooperation Society
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• v.7 no.6
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• pp.1034-1039
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• 2006

The purpose of this paper is to evaluate the structural integrity of the Class 1500 Bellows Seal 3 inch globe valve classified as seismic category IIA. The finite element analysis program, ANSYS, Version 10.0, is used to perform both a modal frequency analysis and an equivalent static stress analysis of the subject valve modeling. The modal frequency analysis results show the fundamental natural frequency is greater than 33 Hz. Therefore the equivalent static stress analysis is performed using the seismic acceleration values. The stresses resulted from various loadings and their combinations are evaluated based on the structural acceptance criteria of the ASME Code. The stresses in the glove valve due to the seismic loadings are within the allowable limits. It is concluded that the globe valve structure is maintaining the structural integrity fur the seismic loading conditions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2001.11a
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• pp.498-498
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• 2001

Seismic qualification of the Main Control Boards for nuclear power plants has been performed with the guideline of AS ME Section III. US NRC Reg. Guide and IEEE 344 code. The analysis model of the Main Control Boards is consist of beam. shell and mass element by using the finite element method. and, at the same time. the excitation forces and other operating loads for each model are encompassed with respect to different loading conditions. As the fundamental frequencies of the structure are found to be less than 33Hz. which is the upper frequency limit of the seismic load, the response spectrum analysis using ANSYS is performed in order to combine the modal stresses within the frequency limit. In order to confirm the structural and functional integrity of the major components, modal analysis theory is adopted to derive the required response spectrum at the component locations. As all the combined stresses obtained from the above procedures are less than allowable stresses and no mechanical or electrical failures are found from the seismic testing, it concludes the Main Control Boards is dynamically qualified for seismic conditions. Although the authors had confirmed the structural and functional integrity of both Main Control Boards and all the component, in this paper only the seismic analysis of the Main Control Board is introduced.

• Journal of the Korean Society of Industry Convergence
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• v.27 no.5
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• pp.1137-1144
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• 2024

This paper is about an electric water pump used in an electric vehicle cooling system. An electric water pump is operated by a BLDC motor compared to a mechanical one, so it operates only as much as necessary, improving fuel efficiency. The use of an electric water pump reduces exhaust gas and has the advantage of being free to install, so it can be applied to automobiles, ships, and aircraft. In order to optimize the production of a BLDC motor used as an electric water pump, FEM and electromagnetic field analysis were performed. The dimensions and materials of the stator and rotor were selected by applying the values obtained through the analysis. In addition, the output characteristics of the motor were analyzed through parameter analysis and shape change through self-equivalent circuit analysis to reduce the outer diameter and increase the torque. The electromagnetic hazard of the PCB board was verified, and power integrity analysis was performed to reduce resonance and noise.