• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3353-3360
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• 2022

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.

• 한국소음진동공학회논문집
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• 제12권11호
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• pp.856-863
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• 2002

Seismic qualification of the main control board(MCB) for the nuclear power plant Ulchin 5 and 6 has been performed with the guideline of ASME Section III and IEEE 344 code. As the size and weight of the MCB are too large and heavy to excite using the excitation table, finite element analysis is used in order to investigate the dynamic behaviors and structural integrity of the MCB. As the fundamental frequencies of the equipment are found to be less than 33 Hz, which is the upper frequency limit for the dynamic analysis, response spectrum analysis using ANSYS is performed in order to combine the modal stresses within the frequency limit. In order to confirm the electrical stability of the major components of the MCB. modal analysis theory has been adopted to derive the required response spectra at the component locations. As the all combined stresses obtained from the above procedures are less than the allowable stresses and no mechanical or electrical failures are found from the seismic testing, the authors can confirm the safety of the nuclear equipment MCB under the given seismic loading conditions.

• 한국태양에너지학회 논문집
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• 제30권3호
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• pp.65-70
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• 2010

As for the present wind power industry, most of the computerization for monitoring and control is based on the traditional development methodology, but it is necessary to improve SCADA system since it has a phenomenon of backlog accumulation in the applicable aspect of back-data as well as in the operational aspect in the future. Especially for a system like offshore wind power where a superintendent cannot reside, it is desirable to operate a remote control system. Therefore, it is essential to establish a monitoring system with appropriate control and monitoring inevitably premised on the integrity and independence of data. As a result, a study was carried out on the modeling of offshore wind power data-centered database. In this paper, a logical data modeling method was proposed and designed to establish the database of offshore wind power. In order for designing the logical data modeling of an offshore wind power system, this study carried out an analysis of design elements for the database of offshore wind power and described considerations and problems as well. Through a comparative analysis of the final database of the newly-designed off-shore wind power system against the existing SCADA System, this study proposed a new direction to bring about progress toward a smart wind power system, showing a possibility of a service-oriented smart wind power system, such as future prediction, hindrance-cause examination and fault analyses, through the database integrating various control signals, geographical information and data about surrounding environments.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 추계학술대회논문집
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• pp.253-253
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• 2008

The main purpose of this study is to identify the vibrational characteristics for the LP blades of Korean standard fossil power plants. Modal tests for the 6 stage blade with boundary condition in which the root of blades are constrained with the disk were conducted, and FE analysis was also did with the same boundary condition. The steady-stress and modal analyses for the coupled bladed-disk system of LP turbine stages were completed. The dynamic analysis and fatigue analysis were followed to diagnose the integrity of LP turbine blades.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 추계학술대회 논문집
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• pp.306-311
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• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.1-11
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• 2015

Feedwater Heaters are important components in a nuclear power plant. As the age of heater increases, the maintenance cost required for continuous operation also increases. Most heaters have the carbon steel shells, tube support plates and flow baffles. The carbon steel is susceptible to flow-accelerated corrosion. This is especially true if the flow has a two-phase mixture of steam and condensate. The wall thinning around the wet steam entrance area of the shell is inevitable during some long term operation. The structural integrity of the feedwater heater shell affects the safe operation of the nuclear power plant. Therefore, it is needed for the thinned shell to be repaired. The maintenance method for preventing failure of the shell should be determined by investigating various factors including the stress distribution of thinned area. The stress analysis of the shell including the steam entrance region is studied in this paper. The results of thinned shell is compared with that of intact shell.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.225-234
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• 2014

Wall-thinned defects caused by accelerated corrosion due to fluid flow in the inner pipe appear in many structures of the secondary systems in nuclear power plants (NPPs) and are a major factor in degrading the integrity of pipes. Wall-thinned defects need to be managed not only when the NPP is under maintenance but also when the NPP is in normal operation. To this end, a test technique was developed in this study to detect such wall-thinned defects based on the temperature difference on the surface of a hot pipe using infrared (IR) thermography and a cooling device. Finite element analysis (FEA) was conducted to examine the tendency and experimental conditions for the cooling experiment. Based on the FEA results, the equipment was configured before the cooling experiment was conducted. The IR camera was then used to detect defects in the inner pipe of the pipe specimen that had artificially induced defects. The IR thermography developed in this study is expected to help resolve the issues related to the limitations of non-destructive inspection techniques that are currently conducted for NPP secondary systems and is expected to be very useful on the NPPs site.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2006년도 학술발표회 논문집
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• pp.56-63
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• 2006

Facilities of 76skV Substation(S/S) play an important role in electric power supply grids. Various power facilities of 765kV S/S might be damaged enormously if a strong earthquake occurs. In an effort to mitigate possible earthquake disasters, KEPRI (Korea Electric Power Research Institute) set forth plans to verify seismic safety of the facilities of 765kV S/S. To accomplish the task, an earthquake monitoring systems is constructed at four 765kV S/S sites(Shin-AnSung, Shin-TaeBaek, Shin-SeoSan and Shin-GaPyung). Data from these earthquake monitoring stations are being transmitted via satellite communication. Currently, KEPRI is operating an earthquake monitoring system in freefield of Shin-SeoSan S/S (NSS) tentatively, Also, the data from NSS is preliminarily analyzed using the horizontal to vertical (H/V) spectrum ratio method. The method of H/V spectrum ratio has been used to infer site amplification without previous knowledge of near surface geology. The results of data analysis shorts good S/N ratio and amplification of 20-25 Hz by site effect. In the near future, the accumulated data is expected to provide a basis for assessing and predicting any damages to integrity of 765kV S/S facilities by earthquakes.