• Nuclear Engineering and Technology
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• 제42권1호
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.870-880
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• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

• Environmental Engineering Research
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• 제7권4호
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• pp.227-237
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• 2002

Chemical vapor deposition (CVD), ion beam sputtering (IBS) and sol-gel method were used to prepare TiO$_2$ thin films for degradation of hazardous organic compounds exemplified by 2-chlorophenol (2-CP). The influence of supporting materials and coating methods on the photocatalytic activity of the TiO$_2$ thin films were also studied. TiO$_2$ thin films were coated onto various supporting materials including steel cloth (SS), copper cloth, quartz glass tube (QGT), and silica gel (SG). Results indicate that SS (37 μm)- TiO$_2$ thin film prepared by IBS method improves the photodegradation of 2-CP. Among all supporting materials studied, SS(37 μm) is found to be the best support.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기기기 및 에너지변환시스템부문
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• pp.50-52
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• 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.

• 한국정보과학회:학술대회논문집
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• 한국정보과학회 2002년도 가을 학술발표논문집 Vol.29 No.2 (1)
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• pp.271-273
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• 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.

• Bulletin of the Korean Chemical Society
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• 제22권11호
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• pp.1183-1191
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• 2001

The metallorganic chemical vapor deposition (MOCVD) method has been used to prepare TiO2 thin films for the degradation of hazardous organic compounds, such as 2-chlorophenol (2-CP). The effect of supporting materials and metal doping on the photocatalytic activity of TiO2 thin films also has been studied. TiO2 thin films were coated onto various supporting materials, including stainless steel cloth(SS), quartz glass tube (QGT), and silica gel (SG). Transition metals, such as Pd(II), Pt(IV), Nd(III) and Fe(III), were doped onto TiO2 thin film. The results indicate that Nd(Ⅲ) doping improves the photodegradation of 2-CP. Among all supporting materials studied, SS(37 ${\mu}m)$ appears to be the best support. An optimal amount of doping material at 1.0 percent (w/w) of TiO2-substrate thin film gives the best photodegration of 2-CP.

• 비파괴검사학회지
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• 제26권2호
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• pp.90-98
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• 2006

원전 증기 발생기 세관의 MRPC probe 신호를 검사하고 평가하기 위해서는 일반적으로 Combo 표준 보정 시험편 신호가 신호 보정을 위해 사용된다 이렇듯 Combo 표준 보정 시험편 신호는 신호 평가에 중요한 영향을 미치지만, probe 상태나 관 주위의 여러 요소에 의해 쉽게 영향을 받기 때문에 결함 평가 요소인 신호의 크기 값과 위상각을 왜곡시킬 수 있다. 따라서 본 연구는 이런 문제점을 극복하기 위해 Combo 표준 보정 시험편의 실제 신호를 모사 신호로 대체하는 가능성을 알아보기 위해 실험을 해 보았다 이를 위해 MRPC probe와 Combo 표준 보정 시험편의 특성을 조사하였으며 계산 수행을 위해 체적 적분 방법으로 계산되는 상용 전자기 해석 프로그램인 VIC-3D를 사용하였고 모사 신호를 생성한 후 실험 신호와 비교를 통해 신호의 정확성을 확인하였다. 마지막으로, 모사 신호를 이용한 결함 평가를 위하여 실제 결함과 가공 결함에 대해 위상각과 크기 값의 항목으로 평가하여 실제 결함 평가자에 의한 결과와 비교하였다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.893-899
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• 2024

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju". However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H₂), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H₂ bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.513-522
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• 2013

The ODSCC detected in the TSP position of Ulchin 3&4 SGs are typical ODSCC of Alloy 600MA tubes. The causative chemical environment is formed by concentration of impurities inside the occluded region formed by the tube surface, egg crate strips, and sludge deposit there. Most cracks are detected at or near the line contacts between the tube surface and the egg crate strips. The region of dense crack population, as defined as between $4^{th}$ and $9^{th}$ TSPs, and near the center of hot leg hemisphere plane, coincided well with the region of preferential sludge deposition as defined by thermal hydraulics calculation using SGAP computer code. The cracks developed homogeneously in a wide range of SGs, so that the number of cracks detected each outage increased very rapidly since the first detection in the $8^{th}$ refueling outage. The root cause assessment focused on investigation of the difference in microstructure and manufacturing residual stress in order to reveal the cause of different susceptibilities to ODSCC among identical six units. The manufacturing residual stress as measured by XRD on OD surface and by split tube method indicated that the high residual stress of Alloy 600MA tube played a critical role in developing ODSCC. The level of residual stress showed substantial variations among the six units depending on details of straightening and OD grinding processes. Youngwang 3&4 tubes are less susceptible to ODSCC than U3 and U4 tubes because semi-continuous coarse chromium carbides are formed along the grain boundary of Y3&4 tubes, while there are finer less continuous chromium carbides in U3 and U4. The different carbide morphology is caused by the difference in cooling rate after mill anneal. There is a possibility that high chromium content in the Y3&4 tubes, still within the allowable range of Alloy 600, has made some contribution to the improved resistance to ODSCC. It is anticipated that ODSCC in Y5&6 SGs will be retarded more considerably than U3 SGs since the manufacturing residual stress in Y5&6 tubes is substantially lower than in U3 tubes, while the microstructure is similar with each other.

• 소성∙가공
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• 제23권7호
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• pp.405-412
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• 2014

As a major type of heat exchanger, the steam generator (SG) produces steam from heat energy of a nuclear power plant reactor. The steam produced by the steam generator flows into a turbine, and plays an important role in electric power generation. The heat transfer tubes in the steam generator consist of approximately 10,000 U-shaped tubes, which perform a structural role and act as thermal boundaries. The heat transfer tubes conduct the thermal energy between the primary coolant (about $320^{\circ}C$, $157kgf/cm^2$) obtained from the reactor and the secondary coolant (about $260^{\circ}C$, $60kgf/cm^2$) as part of the secondary system. Recently, the heat transfer tubes in the steam generator of the pressurized water reactor (PWR) are primarily produced from Alloy 600 and Alloy 690 seamless tubes. As a pilot study to find process parameters for the cold U-bending process using rotary draw bending, numerical and experimental investigations were conducted to produce U-shaped tubes from long straight SUS304L seamless tubes. 3D finite element simulations were run using ABAQUS Explicit with consideration of the elastic recovery. The process parameters studied were the angular speed, the operation period and the bending angle. Experimental verifications were conducted to insure the suitability of the final U-shaped configurations with respect to both ovality and wall thickness.