• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 춘계학술대회
• /
• pp.2047-2052
• /
• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2001년도 제34회 추계학술대회 개최
• /
• pp.59-62
• /
• 2001

Flow induced vibration in steam generators has caused dynamic interactions between tubes and contacting materials resulting in fretting wear . Series of experiments have been performed to examine the wear properties of Inconel 690 steam generator tubes in various environmental conditions. For the present study, the test rig was designed to examine the fretting wear and rolling wear properties in high temperature(room temperature - 290。C) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of the steam generator tube materials. To investigate the wear mechanism of material, the worn was observed using scanning electron microscopy. The weight loss increase at higher test temperature was caused by the decrease of water viscosity and the mechanical property change of tube material. The mechanical property changes of steam generator tube material, such as decrease of hardness or yield stress in the high temperature tests. From the SEM observation of worn surfaces, the severe wear scars were observed in specimens tested at the higher temperature.

• 조명전기설비학회논문지
• /
• 제12권2호
• /
• pp.69-76
• /
• 1998

본 논문에서는 배압터빈과 추기복수터빈으로 이루어진 열병합 발전플랜트의 최전운전을 결정할 수 있는 새로운 알고리즘을 제시한다. 제시한 알고리즘은 플랜트가 운전중에 직접 온 라인으로 취할 수 있는 증가량만을 파라메타로 하여 보일러와 터빈-발전기의 최적부하를 결정할 수 있다. 본 알고리즘은 비선형 경비함수와 해당 제한사항들로 이루어져 있으며 실제 운전중인 열병합 발전플랜트와 비교 시뮬레이션을 실시한 결과 만족할만한 결과를 얻었다. 즉 실제 운전 데이터와 비교해본 결과 공정의 증기 부하량에 따라 1.2∼4.5[%]의 운전경비 절감효과를 얻을 수 있었다. 또한 본 알고리즘은 필요한 입력 데이터를 공정으로부터 쉽게 온 라인으로 취할 수 있어 프로세스 컴퓨터로 용이하게 구현할 수 있다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.551-556
• /
• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2008년도 제39회 하계학술대회
• /
• pp.1663-1664
• /
• 2008

Many years ago, most of thermal power plants built in this country were of subcritical pressure, of medium or small size, of constant pressure operations and of drum type steam generators with circulation type boilers. But, nowadays almost all of them were of high efficiency, of supercritical pressure, of big capacity, of sliding pressure operations, and of once through type steam generators. Presently built once through boilers introduce turbine bypass systems to variable pressure operation which eliminates unexpected materials in boiler tube during startup, minimizes fuel loss by short startup period and eventually improve both total efficiency and power system stability.

• 한국지능시스템학회:학술대회논문집
• /
• 한국퍼지및지능시스템학회 1993년도 Fifth International Fuzzy Systems Association World Congress 93
• /
• pp.1070-1073
• /
• 1993

A Fuzzy Logic Controller for handing the swell/shrink problems of nuclear steam generators is designed, implemented and tested on the compact nuclear simulator at Korea Atomic Energy Research Institute. Its performance is found to be better than of the PI controller originally being used. In terms of the total variations for the control actions and for the flow error curve, the ones by the fuzzy controller are found to be less than one third of those by the PI controller.

• Nuclear Engineering and Technology
• /
• 제34권4호
• /
• pp.358-369
• /
• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• Nuclear Engineering and Technology
• /
• 제35권2호
• /
• pp.121-132
• /
• 2003

To reduce the construction cost and enhance the safety of sodium cooled liquid metal reactors, various kinds of new design concepts were evaluated using the KALIMER operation condition. The required equipment sizes were set for plant electricity output to be similar to that of KALIMER. The evaluations were made focusing on the plant performance and implementation practicality. Each design concept was evaluated for the concept itself and design impacts to interfacing systems. Through the evaluation of the concepts, it was found that the most favorable design concept is the integrated steam generator with forced convection using lead bismuth as the intermediate heat transfer fluid between the primary sodium tube and feed water/steam tube in the steam generator.

• Wind and Structures
• /
• 제23권2호
• /
• pp.157-169
• /
• 2016

The structural integrity of tube bundles represents a major concern when dealing with high risk industries, such as nuclear steam generators, where the rupture of a tube or tubes will lead to the undesired mixing of the primary and secondary fluids. Flow-induced vibration is one of the major concerns that could compromise the structural integrity. The vibration is caused by fluid flow excitation. While there are several excitation mechanisms that could contribute to these vibrations, fluidelastic instability is generally regarded as the most severe. When this mechanism prevails, it could cause serious damage to tube arrays in a very short period of time. The tubes are therefore stiffened by means of supports to avoid these vibrations. To accommodate the thermal expansion of the tube, as well as to facilitate the installation of these tube bundles, clearances are allowed between the tubes and their supports. Progressive tube wear and chemical cleaning gradually increases the clearances between the tubes and their supports, which can lead to more frequent and severe tube/support impact and rubbing. These increased impacts can lead to tube damage due to fatigue and/or wear at the support locations. This paper presents simulations of a loosely supported multi-span U-bend tube subjected to turbulence and fluidelastic instability forces. The mathematical model for the loosely-supported tubes and the fluidelastic instability model is presented. The model is then utilized to simulate the nonlinear response of a U-bend tube with flat bar supports subjected to cross-flow. The effect of the support clearance as well as the support offset are investigated. Special attention is given to the tube/support interaction parameters that affect wear, such as impact and normal work rate.

• Corrosion Science and Technology
• /
• 제15권2호
• /
• pp.84-91
• /
• 2016

The transport and deposition of corrosion products in pressurized water nuclear reactor (PWR) steam generators have led to corrosion (SCC, denting etc.) problems. Lancing, mechanical cleaning and chemical cleaning have been used to reduce these problems. The methods of lancing and mechanical cleaning have limitations in removing corrosion products due to the structure of steam generator tubes. But high temperature chemical cleaning (HTCC) with EDTA is the most effective method to remove corrosion products regardless of the structure. However, EDTA in chemical cleaning aqueous solution and chemical cleaning environments affects the integrity of materials used in steam generators. The nuclear power plants have to perform the pre-test (also called as qualification test (QT)) that confirms the effect on the integrity of materials after HTCC. This is one of the series studies that assess the effect, and this study determines the effects of 20 % EDTA aqueous solution on defective tubes in high temperature chemical cleaning environments. The depth and magnitude of defects in steam generator (SG) tubes were measured by eddy current test (ECT) signals. Surface analysis and magnitude of defects were performed by using SEM/EDS. Corrosion rate was assessed by weight loss of specimens. The ECT signals (potential and depth %) of defective tubes increased marginally. But the lengths of defects, oxides on the surface and weights of specimens did not change. The average corrosion rate of standard corrosion specimens was negligible. But the surfaces on specimens showed traces of etching. The depth of etching showed a range on the nanometer. After comprehensive evaluation of all the results, it is concluded that 20 % EDTA aqueous solution in high temperature chemical cleaning environments does not have a negative effect on defective tubes.