• Nuclear Engineering and Technology
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• 제35권6호
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• pp.566-580
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• 2003

To resolve the concern of the SWR possibility in LMR and improve the economic feature of LMR, relative performance of various SG designs using a double tube bundle configuration is evaluated and a new SG design concept is proposed. The new steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially divided into two. It prevents the occurrence of sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of the SG with a double tube bundle for actual use in an LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles. The evaluation revealed the relative advantages and disadvantages of the configurations and the new SG design concept performs good and can be actually used in an LMR plant.

• 한국소음진동공학회논문집
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• 제26권2호
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• pp.203-209
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• 2016

According to the wear detection history for the steam generator tubes in the nuclear power plant, the outer tubes inside the steam generator have more problems on the flow-induced vibration than inner tubes. Many researchers and engineers have used a specified added mass coefficient for a given tube array during the design stage of the steam generator even though the coefficient is not constant for entire tube in cylindrical shell. The aim of this study is to find out the distribution of added mass coefficients for each tube along the radial location. When numbers of tubes inside a cylindrical shell are increased, values of added mass coefficients are also increased. Added mass coefficients at outer tubes are less than those of inner tubes and they are decreased with increasing the gap between the outermost tube and the cylindrical shell. It also turns out when the gap between the outermost tube and the cylindrical shell approaches infinite value, the added mass coefficient converges to an asymptotic value of given tube array in a free fluid field.

• 비파괴검사학회지
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• 제33권6호
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• pp.511-518
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• 2013

Steam Generator (SG) tube, as a barrier isolating the primary coolant system from the secondary side of nuclear power plants (NPP), must maintain the structural integrity for the public safety and their efficient power generation. So, SG tubes are subject to the periodic examination and the repairs if needed so that any defective tubes are not in service. Recently, corrosion related degradations were detected in the tubes of the domestic OPR-1000 NPP, as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). According to the studies on the factors causing the heat fouling as well as developing corrosion cracking, densely scaled deposits on the secondary side of the SG tubes are mainly known to be problematic causing the adverse impacts against the soundness of the SG tubes [1]. Therefore, the processes of various cleaning methods efficiently to dissolve and remove the deposits have been applied as well as it is imperative to maintain the structural integrity of the tubes after exposing to the cleaning agent. So qualification test (QT) should be carried out to assess the perfection of the chemical cleaning and QT is to apply the processes and to do ECT. In this paper, the chemical cleaning processes to dissolve and remove the scaled deposits are introduced and results of ECT on the artificial crack specimens to determine the effectiveness of those processes are represented.

• 비파괴검사학회지
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• 제30권1호
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• pp.13-19
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• 2010

본 논문에서는 원전SG세관 결함 크기 추정을 위한 새로운 구조의 추정시스템에 대한 연구를 수행한다. 기존의 연구에서는 결함 크기를 추정하기 위하여 각각의 결함 형태별로 결함크기추정시스템을 설계하였다. 이와 같은 경우, 추정시스템의 구조가 복잡해지고 결함 크기 추정 이전에 수행하는 결함형태분류기의 정확성이 떨어질 경우 결함 크기 추정 성능도 결과적으로 악화될 수밖에 없다. 이에 본 논문에서는 결함 형태 분류 과정을 필요로 하지 않는 결함크기추정시스템의 성능을 분석하고 이를 향상시키기 위한 방안을 연구하였다. 기존의 추정시스템은 각각의 결함 형태별로 특화된 추정기를 사용하기 때문에 추정 성능이 훨씬 뛰어날 것으로 예상되었지만, 실험 결과 두 추정시스템의 성능 차이는 그리 크지 않다는 것을 알 수 있었다. 따라서 결함형태분류기의 정확성이 완벽하지 않을 경우, 본 논문에서 제안한 구조의 추정기가 효과적으로 사용될 수 있을 것으로 기대된다.

• 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.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 1998년도 춘계 학술대회논문집
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• pp.75-81
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• 1998

A new advanced integral reactor of 330 MWt capacity named SMART(System-integrated Modular Advanced ReacTor) is currently under development at KAERI(Korea Atomic Energy Research Institute). One of the major design features of the integral reactor is locating the steam generators(SG) inside reactor vessel and eliminating the possibility of LB LOCA(large Break Loss of Coolant Accident). Orifices are fitted at the low part of steam generator cassette to stabilize and balance coolant flow distribution in the MCP (Main Circulation Pump) pressure header. A sensitivity analysis is conducted to determine the optimum orifice size using computer code 'CFX'.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2008년도 춘계학술발표대회 개요집
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• pp.10-10
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• 2008

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.61-66
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• 2011

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

• 전자공학회논문지S
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• 제35S권2호
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• pp.137-145
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• 1998

In this paper, we propose a tube-hole center detection vision algorithm verifying the position of a tele-controlled robot and providing visual information for increasing reliability and efficiency in the diagnosis of steam generator (SG) tubes in nuclear power plant. A tele-controlled robot plays a role in carrying the probe used in inspecting the integrity of SG tubes. Thus accurately locating a tele-controlled robot on the desired tube-hole center is important issue for reliability of inspection. To do this work, we have to find the tube-hole center locations from the input image. At first, we apply the three-class segmentation method modified for this application. WE extract minimum bounding rectangles (MBRs) in the theresholded binary image. Second, for discriminating between MBR by tube and MBR by noise, we introduce the MBR rejection rules as knowledge-based rule set. MBRs are divided into the very dark region MBRs and the very bright region MBRs. In order to describe the region of complete tube-hole, the MBRs need a process of pairing each other. We then can find the tube-hole center from the paired MBR. For more accurately finding the tube-hole center in several sequential images, the centers of some frames need to be averaged. We tested the performance of our method using hundreds of real images.

• Corrosion Science and Technology
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• 제10권4호
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• pp.118-124
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• 2011

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).