• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.4
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• pp.339-345
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• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.

• Proceedings of the KWS Conference
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• 2009.11a
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• pp.62-62
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• 2009

석탄화력발전소의 CO2배출량 감소와 고효율, 대용량화로 인해 초초임계압(USC:Ultra Super Critical) 화력발전소의 건설이 증가하고 있다. USC 발전소는 효율향상을 위한 증기온도와 압력의 상승 때문에 보일러 고온고압부에 기존의 소재에 비해 고온강도와 내산화성의 재료물성이 향상된 신소재 적용이 불가피하다. 특히 사용된 신소재 중에서 보일러 본체를 구성하는 수냉벽관(Water wall), 과열기와, 재열기용 튜브 및 후육부인 헤더와 배관재로 기존의 2.25Cr-1Mo강을 개량한 2.25Cr-1.6W계 내열강이 적용되고 있다. 2.25Cr-1.6W강은 SMI와 MHI가 공동개발한 소재로 1995년 튜브제품이, 1999년에 단조, 파이프재, 플레이트제품이 ASME code case로 등재되었고, 2009년 ASME code case 2199-4로 개정되어 사용 중이다. 이 소재는 2.25Cr-1Mo강에 고온강도 개선을 위해 석출강화효과가 있는 V과 Nb을 첨가하였고, 탄화물의 열적안정성과 고용강화효과 증대를 위해 W을 첨가하였다. 그리고 제작성과 용접성 및 재료의 인성 향상을 위해 B첨가와 C함량을 낮추었다. 합금성분의 첨가와 조정에 의해 고온강도는 개선되었지만, 보일러 설치 및 보수를 위한 용접과정에서 용접금속과 CGHAZ(Coarse Grain HAZ)에서 용접균열이 발생하였다. 대부분의 용접균열은 용접결함이나 고온 혹은 저온균열이 아닌 2.25Cr-1.6W계강의 강도 개선을 위해 첨가한 V과 Nb이 용접후열처리 도중 입내에 MX형태의 미세석출로 입내를 강화시킴으로서 발생한 재열균열 민감성 증대에 기인된 것으로 판단된다. 이에 본 연구에서 용접 및 후열처리 과정에서 용접금속과 HAZ에서 발생하는 용접금속의 응력분포를 전산해석을 통해 확인하고 실제 후육파이프 용접부에서 잔류응력을 측정해 비교하였다. 용접부 응력분포는 SYSWELD 프로그램을 사용해 해석을 수행하였고, 발전소 실배관재의 용접부 응력측정은 수평부 측정이 용이하도록 지그를 부착한 Potable 잔류응력측정기를 사용해 Hole Drilling Method(HDM)를 적용하여 잔류응력을 측정하였다. 해석 결과 CGHAZ부위의 잔류응력이 용접금속과 기타 부위에 비해 높은 응력분포를 나타냈으며, 이는 CGHAZ와 용접용융선 부근에서 균열이 발생하는 실제값과 일치하는 결과를 보였다. 실제 배관재 용접부에서 측정한 잔류응력값은 항복응력의 약 50% 이하 응력값을 나타냈다. 배관 구조에 기인한 시스템응력의 영향을 제거하기 위해 배관재 용접부를 중심으로 양끝단을 절단 후 용접부에서 측정한 응력은 항복응력 대비 25%수준의 낮은값을 보였다. 그러나 배관재가 장기간 고온환경에 노출되었고 용접금속 내부의 균열이 발생한 상태에서 측정하였기 때문에 용접잔류응력은 상당부분 해소되어 상대적으로 낮은 응력값이 얻어진 것으로 판단된다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.8
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• pp.783-789
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• 2012

In this study, turbulent flow and heat transfer characteristics in a helically coiled tube have been numerically investigated. Helically coiled tubes are commonly used in heat exchange systems to enhance the heat transfer rate. Accordingly, they have been widely studied experimentally; however, most studies have focused on the pressure drop and heat transfer correlations. The centrifugal force caused by a helical tube increases the wall shear stress and heat transfer rate on the outer side of the helical tube while decreasing those on the inner side of the tube. Therefore, this study quantitatively shows the variation of the local Nusselt number and friction factor along the circumference at the wall of a helical tube by varying the coil diameter and Reynolds number. It is seen that the local heat transfer rate and wall shear stress greatly decrease near the inner side of the tube, which can affect the safety of the tube materials. Moreover, this study verifies the previous experimental correlations for the friction factor and Nusselt number, and it shows that the correlation between the two in a straight tube can be applied to a helical tube. It is expected that the results of this study can be used as important data for the safety evaluation of heat exchangers and steam generators.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.6
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• pp.504-509
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• 2016

Defects in dissimilar metal welds are reported to be on the increase during the operating lifespan and aging of nuclear power plants. In Korea, reported cases of defects due to dissimilar metal welds include the drain nozzle of a steam generator and RCS hot tube sampling nozzles. Therefore, there is an urgent need to develop a reliable automated nondestructive inspection technique and a system for the inspection of dissimilar metal welds of small diameter nozzles in a high radiation area of a nuclear power plant. In this study, to ensure effective defect inspection of small diameter nozzles (RCS high-temperature tube sampling nozzle) of a nuclear power plant, three different methods were developed. These include: (1) optimum inspection probe design by beam simulation, (2) multi-directions UT optimum inspection technique for the inspection of small diameters of different welded parts, and (3) remote control automatic inspection system. The developed technique and systems have been verified to be suitable for use in the inspection of defects in smaller diameter nozzles in nuclear power plants.

• Korean Chemical Engineering Research
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• v.47 no.6
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• pp.795-800
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• 2009

In this study, the $CO_2$ removal characteristics of the Vortex tube type absorbtion apparatus were investigated to enhance the compactness of $CO_2$ absorption process and to reduce the amount of absorbing solution of the $CO_2$ separation process. The Vortex tube with the diameter of 17 mm and the length of 250mm was introduced in the experimental apparatus to treat $20Nm^3/hr$ of $CO_2$ containing flue gas. The flue gases for experiments containing 11~13 vol% of $CO_2$ were supplied from the coal-firing CFBC power plant with 12 ton/hr of steam producing capacity. The mixed solutions of 20 wt% of MEA as base solution with the adding solutions like HMDA, AMP and KOH were used as absorbents. The experiments were executed under the various conditions like the absorbing solution concentrations in the range of 20 to 50 wt%, the flow rate of $CO_2$ containing flue gases in the range of 6 to $15Nm^3/hr$ and the flow rate of absorbing solution in the range of 1.0 to 3.0 l/min. As a results, the $CO_2$ removal efficiency of mixed absorbent of 20 wt% of MEA with HMDA was remarkable. From this study, we concluded that the efficient separation of $CO_2$ from flue gases using the features of the Vortex tube type absorbing unit for gas/liquid contact and the separation of gas/liquid be possible. But more works are needed to increase the $CO_2$ removal efficiency of Vortex tube process.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.05a
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• pp.312-317
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the PIAT (Program for Integrity Assessment of Steam Generator Tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-vibration bars such as vertical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. 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), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.3
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• pp.432-439
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• 2003

Inconel 690 for nuclear steam generator tube has more Chromium than the conventionally used Inconel 600 in order to increase the corrosion resistance. To evaluate the tribological characteristics of Inconel 690 under fretting condition the fretting tests were carried out in air and elevated temperature water. Fretting tests of the cross-cylinder type were done under various vibrating amplitudes and applied normal loads in order to measure the friction forces and wear volumes. From the results of fretting wear tests. the wear of Inconel 690 can be predictable using the work rate model. The amounts of friction forces were proportional to relative movement between two fretting surfaces. The friction coefficients were decreased as increasing the normal loads and deceasing the vibrating amplitudes. Depending on fretting environment, distinctively different wear mechanisms and often drastically different wear rates can occur It was found that the fretting wearfactors in air and water at 2$0^{\circ}C$, 5$0^{\circ}C$, and 8$0^{\circ}C$ were 7.38 $\times$ $10^{-13}$$Pa^{-1}$, 2.12 $\times$$10^{-13}$$Pa^{-1}$, 3.34$\times$$10^{-13}$$Pa^{-1}$and 5.21$\times$$10^{-13}$$Pa^{-1}$, respectively flexibility to model response data with multiple local extreme. In this study, metamodeling techniques are adopted to carry out the shape optimization of a funnel of Cathode Ray Tube, which finds the shape minimizing the local maximum principal stress. Optimum designs using two metamodels are compared and proper metamodel is recommended based on this research.