• Proceedings of the KIEE Conference
• /
• 대한전기학회 2005년도 추계학술대회 논문집 전기기기 및 에너지변환시스템부문
• /
• pp.50-52
• /
• 2005

In this paper, the analysis of the eddy current testing(ECT) signals under thc Influence of the ferromagnetic support plate was performed in steam generator(SG) tube of nuclear power plant. In order to remove the influence of the ferromagnetic support plate, a multi-frequency ECT was used. The models which was established for the analysis of the signals is calculated using numerical analysis of finite element method. Through the result of numerical analysis, improved signals is acquired considering the influence of the ferromagnetic support plate using mixing of multi-frequency This paper is presented the residual errors and the phase changes for analysis of the defect signals which should be considered when conducting a ECT using multi-frequency.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 한국소음진동공학회 1997년도 추계학술대회논문집; 한국과학기술회관; 6 Nov. 1997
• /
• pp.96-100
• /
• 1997

A design modification was made on the 9-th stage wheel dovetail of a high-intermodiate pressure (HIP) turbine rotor for a fossil power plant that necessitates the use of new long-shank buckets for the row. A bucket vibration test is necessary to verify that the new 9-th stage buckets have adequate frequency margin from a nozzle passing frequency when running at speed. A finite element analysis (FEA) has been performed using a commercial S/W to approximately estimate bucket natural frequencies, and thus to help the vibration test. A row of the new buckets has been assembled on the HIP rotor for the vibration tests using dynamic balancing facilities. The tests have been done during deceleration run with air excitation. The test results are compared with the calculation using our empirical formula, and show that the modified design meets the frequency-margin requirements.

• Proceedings of the Korean Information Science Society Conference
• /
• 한국정보과학회 2002년도 가을 학술발표논문집 Vol.29 No.2 (1)
• /
• pp.271-273
• /
• 2002

증기발생기(SG : Steam Generator)는 다수의 세관으로 구성되어 원자로에서 발생한 열을 이용하여 발전기 터빈을 구동시키는 원동력인 증기를 생성해 주는 기능을 하는 원자력발전소의 핵심 설비이다. 증기발생기 세관의 건전성을 확보하기 위해 매주기 계획예방정비, 즉 가동중 검사마다 정기적인 와전류 검사를 수행하고, 검사결과에 따라 전열관 보수 등과 같은 제반 조치를 취하고 있다. 현재 검사데이터 DB 구축은 일부 발전소에 개발되어 운영 중에 있고, 세관 DB와는 별도로 통계정보만을 관리하는 증기발생기 성능관리시스템이 운영되고 있으며, 또한 각 발전소마다 수질을 계측하여 수화학 성분을 감시하는 수질관리시스템이 운용되고 있다. 이러한 이원화된 DB 및 시스템을 통합하고 연계하여 전 원전의 증기발생기를 종합적으로 관리 할 수 있는 시스템의 필요성이 대두되었다. 따라서 본 논문에서는 현장에 보관되어 있는 모든 세관 검사데이터를 취득하여 대용량 데이터베이스를 설계 및 구축하고 이기종의 분산된 수질관리시스템 DB를 연계하여, 증기발생기의 설계/제작부터 검사결과 Mapping, 추이 분석을 통한 수명 평가에 이르는 전 과정을 통합 관리할 수 있는 시스템을 설계하고 그 구현방안을 제시한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• 제15권1호
• /
• pp.8-17
• /
• 2019

Perforated plates are used for the steam generator tube-sheet and the Reactor Vessel Closure Head in the Nuclear Power Plant. The ASME code, Section III Appendix A-8000, addresses the analysis of perforated plates, however, this analysis is only limited to the flat plate with a triangular perforation pattern. Based on the concept of the effective elastic constants, simulation of flat and spherical perforated plates and their equivalent solid plates were carried out using Finite Element Analysis (FEA). The isotropic material properties of the perforated plate were replaced with anisotropic material properties of the equivalent solid plate and subjected to the same loading conditions. The generated curves of effective elastic constants vs ligament efficiency for the flat perforated plate were in agreement with the design curve provided by ASME code. With this result, a plate with spherical curvature having perforations can be conveniently analyzed with equivalent elastic modulus and equivalent Poisson's ratio.

• Nuclear Engineering and Technology
• /
• 제39권2호
• /
• pp.133-142
• /
• 2007

The Electric Power Research Institute (EPRI) demonstrated a method for monitoring the performance of instrument channels in Topical Report (TR) 104965, 'On-Line Monitoring of Instrument Channel Performance.' This paper presents the results of several models originally developed by EPRI to monitor three nuclear plant sensor sets: Pressurizer Level, Reactor Protection System (RPS) Loop A, and Reactor Coolant System (RCS) Loop A Steam Generator (SG) Level. The sensor sets investigated include one redundant sensor model and two non-redundant sensor models. Each model employs an Auto-Associative Kernel Regression (AAKR) model architecture to predict correct sensor behavior. Performance of each of the developed models is evaluated using four metrics: accuracy, auto-sensitivity, cross-sensitivity, and newly developed Error Uncertainty Limit Monitoring (EULM) detectability. The uncertainty estimate for each model is also calculated through two methods: analytic formulas and Monte Carlo estimation. The uncertainty estimates are verified by calculating confidence interval coverages to assure that 95% of the measured data fall within the confidence intervals. The model performance evaluation identified the Pressurizer Level model as acceptable for on-line monitoring (OLM) implementation. The other two models, RPS Loop A and RCS Loop A SG Level, highlight two common problems that occur in model development and evaluation, namely faulty data and poor signal selection

• Nuclear Engineering and Technology
• /
• 제39권2호
• /
• pp.103-110
• /
• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Nuclear Engineering and Technology
• /
• 제19권3호
• /
• pp.198-204
• /
• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Journal of Surface Science and Engineering
• /
• 제49권4호
• /
• pp.389-394
• /
• 2016

The operation method of nuclear power plants is currently changing to high burn-up and long period that can enhance economics and efficiency of the plant. Since nuclear plant operation environment has been becoming severe, the amount of absorbed hydrogen also has increased. Absorbed hydrogen can be fatal securing safety of nuclear fuel cladding in case of Loss of Coolant Accidents(LOCA). In order to examine the impact of hydride on high-temperature oxidation, high-temperature oxidation experiment was performed on normal Zry-4 cladding and on Zry-4 cladding where hydrogen is charged in air pressure steam atmosphere under the $950^{\circ}C$ and $1000^{\circ}C$. According to the results, while oxidation acceleration due to charged hydrogen was not observed prior to breakaway oxidation creation, oxidation began to accelerate in cladding where hydrogens charged as soon as the breakaway oxidation started. If so much hydrogen are charged in the cladding, equiaxial monoclinic phase to unstable of stress is formed and it is presumed that oxidation is accelerated because nearby stress caused a crack in equiaxial phase, and that makes corrosion resistance decline sharply.

• Proceedings of the KSME Conference
• /
• 대한기계학회 2007년도 춘계학술대회B
• /
• pp.3458-3463
• /
• 2007

Leakage would happen because of the damage of high temperature and high-pressure valve in nuclear power plant. condition based prevention maintenance is essential by using the suitable method based on local condition. Energy loss prevention can prevent from an accurate test, Local actually and ability. The methods of test for high energy fluid leakage at present are analysis of ${\Delta}$T, AE(Acoustic Emission) analysis, and thermal image. The result for test of AC (Main steam) system in YNG 2 Unit reveals that the AE occurred clearly in leakage situation, but thermal image didn't occur. It is identified that leakage is occurred when the orifice located front and back of valve operates. It shows that making a impatient judgment by using the single method if it is leakage is containing uncertainty. So I think that using the Multi-Measuring method is more sound judgment than Single-Measuring method.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 한국소음진동공학회 2009년도 추계학술대회 논문집
• /
• pp.716-722
• /
• 2009

4000MW급 유연탄발전소는 기존 500MW급 4기의 발전설비 운영조건에서 지역공동체는 긍정적이었으나, 500MW급의 증설 호기가 늘어나면서 지역 공동체는 발전소 증설과 관련하여 기존 동쪽 부지 경계지역의 소음환경에 대하여 부정적인 문제를 제기하였다. 2000MW급 증설이 진행되어온 4000MW급 발전소의 다양한 소음원에 의한 소음파워 증대와 증가방출 과도소음 에 대하여 고저가 다양한 주변지역에 미치는 영향음 평가하고자 3차원 모델링기법을 적용한다. 기존호기를 비롯한 전체의 발전소 전면부 소음원응 비교 시험결과 증가방출 소음 파워가 기존 설계의 변경 제작되어 기존호기에 비해 20~30dBA 높아져 동시 방출사 주변지역에 과도하게 영향을 미치는 상태였으며 500MW급 4기를 추가 증설에 따라 일상소음원에 의한 소음 영향도 증설전 대비 2~3dBA 증가된 것으로 평가되었다. 따라서 증설호기의 과도 소음원 제거 방안으로 대가 방출증가가 회수되도록 증설호기의 증가방출설비를 개선하였고, 증설에 의하여 증가된 소음 영향음 저감하고자 기존 경계지역에 미치는 소음원음 고려하여 방음벽음 설치한 결과 47 ~ 49dBA를 나타내고 인근 주거건물의 전면부 소음은 43dBA이었다.