• 비파괴검사학회지
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• 제33권6호
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• pp.519-523
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• 2013

A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.397-408
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• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3140-3153
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• 2022

A numerical feasibility study was conducted to investigate the thermal-hydraulic characteristics of a steam generator with corrugated plates for a small modular reactor. Accordingly, a one-dimensional thermal-hydraulic analysis code was developed based on the existing state-of-the-art thermal-hydraulic models and correlations for corrugated plate heat exchangers. Subsequently, the pressure loss, heat transfer, and instability characteristics of the steam generator with corrugated plates were investigated according to the chevron angle and mass flux. Additionally, the characteristics of rectangular and disk-type corrugated plate steam generators with equivalent heat transfer areas were analyzed. The steam generator with disk-type corrugated plates exhibited better performance in terms of pressure loss and heat transfer rate than the rectangular type. In addition, when the mass flux decreased from the onset of boiling points, reverse gradients of the total pressure change were observed in both types. Thus, it was confirmed that Ledinegg instability could occur in the steam generator with corrugated plates. However, it was dependent on the chevron angle, and the optimal chevron angle to minimize instability was 45° under the conditions of the present analysis.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1304-1317
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• 2020

The tubes in a heat exchanger, such as a steam generator (SG), are subjected to crossflow, and interaction between tubes and supports can happen, which can cause fretting wear of tubes. Although many experiments and models have been established, some detailed mechanisms are still not sufficiently clear. In this work, more attention is paid to obtain the regulation of impact and sliding in the complex process and many factors, such as excitation forces and clearances. The responses and contact forces were analyzed to obtain clear understanding of the influences of these factors. Room temperature tests in the air were established. The results show that the effect of clearance on the normal work rate is not monotonous and instead has two peaks. The force ratio can influence the normal work rate by changing the distribution of contact angles, which can result in higher sliding in the contact process. Fretting wear tests are conducted, and the wear surfaces are analyzed by a scanning electron microscope (SEM) and energy dispersive X-ray spectrometer (EDX). The results of this work can serve as a reference for impactsliding contact analysis between AVBs and tubes in steam generators.

• 한국전산유체공학회지
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• 제20권1호
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Nuclear Engineering and Technology
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• 제23권3호
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• pp.299-305
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• 1991

복수기를 통한 해수유입은 증기발생기내에 부식환경을 조성시키게 한다. 이론적 분석을 통하여 복수기누설시에 해수증의 불순물인 염소가 2차계통내에 누적되는 경향을 영광원전을 모델로하여 평가하였다. 분석결과 해수누설시에 고농도의 염소가 증기발생기내에 누적되는 것으로 나타났으나, 이는 증기 발생기내의 수질을 산성분위 기로 조성시킬 것으로 판단되었다. 복수기의 최대허용 설계누설(0.5 gpm)시에는 증기발생기 취출수량을 최대로 늘리고, 복수기정화계통을 가동하더라도 증기발생기에 2.3 ppm 및 복수기집수정에 0.6 ppm의 염소가 누적되는 것으로 나타났다. 또한 증기발생기에서의 염소농축계수는 아래와 같이 전적으로 취출수량 및 정화계통효율에만 의존하는 것으로 나타났으며,（equation omitted）취출수 및 정화계통은 2차계통내의 불순물을 제거하는데 효과적인 것으로 평가되었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 추계학술대회 논문집 학회본부 A
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• pp.273-275
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• 1999

This paper presents a dynamic equivalencing method in large electric power system for stability analysis. This method of modeling simplified equivalents for parts of the network outside the study area is to evaluate the stability of a study area modeled in detail. Generators are closely coupled in an electrical sense tend to swing together in groups during disturbances, and this behavior can be exploited to reduce the size of the power system model. The characteristics of generators swing together are referred to as coherency Coherency groups whose generators state trajectory are similar to the other generators state trajectory in the same coherency group by a certain disturbance. In this paper, procedures for forming dynamic equivalents of control devices of coherency-based generating units are proposed and the aggregation of the control devices such as excitation system and governor-turbine system is accomplished by this method. This method can deal with the aggregation of the same type of control devices and combination of hydro and steam unit or the many types of excitation systems. etc. This method is shown to be efficient in reducing the number of control device of generating units with small error in the study group by result of case study presented latter part of this paper.

• 한국기계가공학회지
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• 제17권5호
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• pp.52-57
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• 2018

Most current ships have adopted on-board diesel generators to produce electricity, but the overall efficiency of equipment is down to about 50% due to thermal losses from operations such as exhaust gas, jacket water cooler, scavenge air cooler, etc. Recently, fuel cells have been highlighted as a promising technology to reduce the effect on the environment and have a higher efficiency. Therefore, this paper suggested a solid oxide fuel cell (SOFC)-gas turbine (GT) using waste heat from a SOFC and SOFC-GT-steam turbine (ST) with Rankine cycle. To compare both configurations, the fuel flow rate, current density, cell voltage, electrical power, and overall efficiency were evaluated at different operating loads. The overall efficiency of both SOFC hybrid systems was higher than the conventional system.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2001년도 추계 학술대회논문집
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• pp.106-111
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• 2001

This paper presents the numerical methodology of ATHOS3 code for thermal hydraulic analysis of Pressurized Water Reactor (PWR) steam generators. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer, and numerical solution scheme. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea Kori Unit-1 nuclear power plant and the computed results are presented.

• 한국수소및신에너지학회논문집
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• 제22권4호
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• pp.546-551
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• 2011

The corrosion on steam generator tubes of the secondary side of pressurized water reactor inhibits heat transfer. One of the most efficient techniques improving the heat transfer performance of a nuclear electric generation is a corrosion control. The environmental parameters mostly affecting corrosion are materials and chemical additives. It seems that no further corrosion occurs in steels with Polyacrylic acid polymer dispersant treatment. Polyacrylic acid forms a protective coating with uniform thickness on metal surface. Polyacrylic treatment appears to be the most convenient way to enhance the thermal performance by the thermal conductivity improvement in steam generators.