• Journal of Mechanical Science and Technology
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• 제18권1호
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• pp.74-81
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• 2004

The KSNP Steam Generators (Youngkwang Unit 3 and 4, Ulchin Unit 3 and 4) have a problem of U-tube fretting wear due to Flow Induced Vibration (FIV). In particular, the wear is localized and concentrated in a small area of upper part of U-bend in the Central Cavity region. The region has some conditions susceptible to the FIV, which are high flow velocity, high void fraction, and long unsupported span. Even though the FIV could be occurred by many mechanisms, the main mechanism would be fluid-elastic instability, or turbulent excitation. To remedy the problem, Eggcrate Flow Distribution Plate (EFDP) was installed in the Central Cavity region or Ulchin Unit 5 and 6 steam generators, so that it reduces the flow velocity in the region to a certain level. However, the cause of the FIV and the effectiveness of the EFDP was not thoroughly studied and checked. In this study, therefore the Stability Ratio (SR), which is the ratio of the actual velocity to the critical velocity, was compared between the value before the installation of EFDP and that after. Also the possibility of fluid-elastic instability of KSNP steam generator and the effectiveness of EFDP were checked based on the ATHOS3 code calculation and the Pettigrew's experimental results. The calculated results were plotted in a fluid-elastic instability criteria-diagram (Pettigrew, 1998, Fig. 9). The plotted result showed that KSNP steam generator with EFDP had the margin of Fluid-Elastic Instability by almost 25％.

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

• 원자력산업
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• 제4권4호통권20호
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• pp.35-41
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• 1984

• 한국지능시스템학회:학술대회논문집
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• 한국퍼지및지능시스템학회 1993년도 Fifth International Fuzzy Systems Association World Congress 93
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• pp.1070-1073
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• 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
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• 제51권5호
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.

• 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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 추계학술대회논문집
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.260-266
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• 2015

Twenty sodium-cooled fast reactors (SFRs) have provided valuable experience in design, licensing, and operation. This paper summarizes the important safety criteria and safety guidelines of intermediate sodium systems, steam generators, decay heat removal systems and associated construction materials and in-service inspection. The safety criteria and guidelines provide a sufficient framework for design and licensing, in particular by new entrants in SFRs.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1999년도 추계 학술발표회 논문집
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• pp.125-130
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• 1999

Corrosion of steam generator tubes is the major issue affecting selection of secondary water chemistry parameters. The objective of secondary side water chemistry control is to minimize corrosion damage and to thereby maximize the reliability and economic performance of the secondary system. To achieve this objective, the water chemistry has to be compatible with all parts of the system including steam generators.(Omitted)

• 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.