• 대한기계학회논문집B
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• 제21권10호
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• pp.1371-1379
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• 1997

The high pressure drop is frequently required in the by-pass line of the pump or of the heat exchanger in power plants. However, cavitation produced by a high pressure drop could damage the pipe and pump blades. Conical orifices are adopted to reduce cavitation due to high pressure drop. The discharge coefficients of conical orifice plates were measured by weighing method in the standard water flow system. The discharge coefficients were larger when the ratios of thickness of orifice edge to throat diameter were larger. The noise generated from a conical orifice due to cavitation was measured with a sound level meter and a hydrophone. With increasing the bore diameter of the orifice, the sound pressure level or the noise level due to cavitation became higher. The noise level was suddenly increased at the inception of cavitation.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2878-2887
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• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• 한국표면공학회지
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• 제57권3호
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• pp.192-200
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• 2024

Cu as a heat exchanger tube is an important component in thermal fluid transfer. However, Cu tubes are exposed to stress in certain environments, leading to stress corrosion cracking (SCC). In this study, the effect of Sn addition on microstructure and corrosion characteristics was examined. The microstructural examination revealed the presence of columnar crystal and a grain refinement due to the addition of Sn. Electrochemical measurements showed that the 5 wt.% NH₃ environment was the most vulnerable environment to Cu corrosion, and the corrosion current density increased as stress increased. The immersion test exhibited the formation of Cu2O and Cu(OH)₂ corrosion product in 3.5 wt.% NaCl and 5 wt.% NH₃ environments, respectively. Results indicated that Sn addition to Cu was an important factor in improving the mechanical strength.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.1-8
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• 2024

In this study, the influence of thermal aging on structural integrity is investigated for Gr. 91 steel. A commercial grade Gr. 91 steel is used for the virgin material, and service-exposed Gr. 91 steel is sampled from a steam pipe of a super critical plant. Time versus creep strain curves are obtained through creep tests with various stress levels at 600 ℃ for the virgin and service-exposed Gr. 91 steels, respectively. Based on the creep test results, the improved Omega model is characterized for describing the total creep strain curve for both Gr. 91 steels. The proposed parameters for creep deformation model are used for predicting the steady-state creep strain rate, creep rupture curve, and stress relaxation. Creep-fatigue damage is evaluated for the intermediate heat exchanger (IHX) in a large-scale sodium test facility of STELLA-2 by using creep deformation model with proposed creep parameters and creep rupture curve for both Gr. 91 steels. Based on the comparison results of creep fatigue damage for the virgin and service-exposed Gr. 91 steels, the thermal aging effect has been shown to be significant.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2332-2342
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• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

• 한국재난관리표준학회지
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• 제3권1호
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• pp.43-50
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• 2010

도로에서 발생하는 사고 중 도로결빙에 의한 미끄러짐 사고비율이 높으며, 특히 상습 결빙 지역인 교량과 터널 입출부에서 발생 빈도가 높게 나타난다. 이렇게 부분적으로 발생되는 결빙 현상을 방지하여 도로운영의 안정성을 높이고자 한다. 지열 융설의 기초 연구를 통해 일반 도로포장 재료인 콘크리트, 아스팔트 두 종류의 재료에 대한 융설시 열전도 양상을 파악하고자 한다. 지열융설 시스템 적용시 포장체 하부 열교환 파이프 매설에 따른 열전도도 연구가 필요하다. 도로 포장과 동일 재질의 시험체를 제작하여 냉동고에서 저온상태의 모형시험을 수행 하였고, 모형시험과 동일 조건의 수치해석을 통해 보다 다양한 융설에 대한 변수를 파악하였다.

• 상하수도학회지
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• 제35권2호
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• pp.113-120
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• 2021

High voltage impulse (HVI) has been gained attention as an alternative technique that could control the CaCO₃ scale problems encountered in water main, pipe, cooling tower and heat exchanger vessels. The aim of this study was to investigate the effect of electric field (E) and contact time (t) of HVI on reduction of Ca²⁺ concentration at two different temperatures of 25℃ and 60℃. A kinetic model on the effect of E and t was investigated too. As the E and t increased, the Ca²⁺ concentration decreased more than that of the control (= no HVI). The Ca²⁺ concentration decreased up to 81% at 15 kV/cm at 60℃, which was nearly 2 times greater than the control. With these experimental data-set of reduction of Ca²⁺ concentration under different E and t, the kinetic model was developed. The relationship between E and t required to reduce the concentration of Ca²⁺ by 30% was modeled at each temperature. The empirical model equations were; E^0.83· t = 60.3 at 25℃ and E^0.08· t = 1.1 at 60℃. These equations state the products of Eⁿ and t is always constant, which means that the required contact time can be reduced in accordance with the increment of E and vice versa.

• 한국결정성장학회지
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• 제22권5호
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• pp.227-232
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• 2012

차세대 원자력발전용 고온 열교환기 소재로 이용될 가능성이 높은 Ni-Cr계 고온합금인 Inconel 617과 Hastelloy X의 표면처리에 따른 FLiNaK(LiF-NaF-KF) 용융염 하에서의 고온물성에 대한 연구를 수행하였다. Inconel 617과 Hastelloy X기판 상에 각각 PVD인 arc discharge 및 sputtering법에 의해 TiAlN 및 $Al_2O_3$ 박막을 코팅 하였다. 이러한 표면처리가 이들 합금의 FLiNaK 용융염 하 고온 안정성에 미치는 영향에 대해 연구하였다. 용융염 하 Ni-Cr계 고온합금의 부식 원리를 이해하기 위해, 용융염 수송 loop에 사용 중 파단된 Inconel 파이프에 대한 미세구조 분석을 수행하였다. 표면처리 된 합금들을 $600^{\circ}C$ 용융염 내에서 열처리 하였으며, 열처리 전후 시편들에 대해 상형성, 미세구조 등 고온 물성 변화를 측정하였다. 연구결과 코팅되지 않은 경우 보다 TiAlN 및 $Al_2O_3$ 박막이 코팅된 소재에서 보다 우수한 고온 안정성을 보여주었다.