• 한국항공우주학회지
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• 제32권4호
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• pp.67-72
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• 2004

경험식을 이용한 1차원 해석에 의하여 30톤급 재생냉각 연소기의 냉각 유로 설계를 수행하였다. 1차원 해석에 의한 벽온도는 3차원 CFD 해석과 비교하여 약 100 K의 온도차이를 보였다. 동일한 냉각성능을 유지하면서 냉각 채널의 최대 폭이 4mm 와 2mm인 두 가지 설계안을 제시하였다. 냉각유체의 압력강하는 20% 증가할 것으로 예측되었다. 열차 폐 코팅과 탄소 침착물의 열저항을 고려한 경우, 최대 벽온도는 700K로 예측되었다. 본 연구에서 제시한 냉각 방법은 용량이 부족한 것으로 판단되는 바 막냉각이 추가적으로 적용되어야 할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.

• 시스템엔지니어링학술지
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• 제13권1호
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• pp.33-39
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• 2017

This paper describes the system engineering approach for the heat transfer analysis of plus7 fuel rod for APR1400 using, a commercial software, ANSYS. The fuel rod is composed of fuel pellets, fill gas, end caps, plenum spring and cladding. The heat is transferred from the pellet outward by conduction through the pellet, fill gas and cladding and further by convection from the cladding surface to the coolant in the flow channel. The goal of this paper is to demonstrate the temperature and heat flux change from the fuel centerline to the cladding surface when having maximum fuel centerline temperature at 100% power. This phenomenon is modelled using the ANSYS FEM code and analyzed for steady state temperature distribution across the fuel pellet and clad and the results were compared to the standard values given in APR1400 SSAR. Specifically the applicability of commercial software in the evaluation of nuclear fuel temperature distribution has been accounted. It is note that special codes have been used for fuel rod mechanical analysis which calculates interrelated effects of temperature, pressure, cladding elastic and plastic behavior, fission gas release, and fuel densification and swelling under the time-varying irradiation conditions. To satisfactorily meet this objective we apply system engineering methodologies to formulate the process and allow for verification and validation of the results acquired. The close proximity of the results obtained validated the accuracy of the FEM analysis of the 2D axisymmetric model and 3D model. This result demonstrated the validity of commercial software instead of proprietary in-house code that is more costly to develop and maintain.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1676-1685
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• 2021

We develop a new high-fidelity multiphysics model to simulate boron chemistry in the porous Chalk River Unidentified Deposit (CRUD) deposits. Heat transfer, capillary flow, solute transport, and chemical reactions are fully coupled. The evaporation of coolant in the deposits is included in governing equations modified by the volume-averaged assumption of wick boiling. The axial offset anomaly (AOA) of the Seabrook nuclear power plant is simulated. The new model reasonably predicts the distributions of temperature, pressure, velocity, volumetric boiling heat density, and chemical concentrations. In the thicker CRUD regions, 60% of the total heat is removed by evaporative heat transfer, causing boron species accumulation. The new model successfully shows the quantitative effect of coolant evaporation on the local distributions of boron. The total amount of boron in the CRUD layer increases by a factor of 1.21 when an evaporation-driven increase of soluble and precipitated boron concentrations is reflected. In addition, the concentrations of B(OH)₃ and LiBO₂ are estimated according to various conditions such as different CRUD thickness and porosity. At the end of the cycle in the AOA case, the total mass of boron incorporated in CRUD deposits of a reference single fuel rod is estimated to be about 0.5 mg.

• 한국유체기계학회 논문집
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• 제18권2호
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• pp.60-66
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• 2015

Conjugate heat transfer analysis was performed to investigate the flow and cooling performance of the high pressure turbine nozzle of gas turbine engine. The CHT code was verified by comparison between CFD results and experimental results of C3X vane. The combination of k-${\omega}$ based SST turbulence model and transition model was used to solve the flow and thermal field of the fluid zone and the material property of CMSX-4 was applied to the solid zone. The turbine nozzle has two internal cooling channels and each channel has a complex cooling configurations, such as the film cooling, jet impingement, pedestal and rib turbulator. The parabolic temperature profile was given to the inlet condition of the nozzle to simulate the combustor exit condition. The flow characteristics were analyzed by comparing with uncooled nozzle vane. The Mach number around the vane increased due to the increase of coolant mass flow flowed in the main flow passage. The maximum cooling effectiveness (91 %) at the vane surface is located in the middle of pressure side which is effected by the film cooling and the rib turbulrator. The region of the minimum cooling effectiveness (44.8 %) was positioned at the leading edge. And the results show that the TBC layer increases the average cooling effectiveness up to 18 %.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.257-264
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• 2013

In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Corrosion Science and Technology
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• 제22권1호
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• pp.44-54
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• 2023

The rupture disk corrosion test (RDCT) method was recently developed to evaluate stress corrosion cracking (SCC) and was found to have great potential for the real-time detection of SCC initiation in a high temperature and pressure environment, simulating the primary water coolant of pressurized water reactors. However, it is difficult to directly measure the stress applied to a disk specimen, which is an essential factor in SCC initiation. In this work, finite element analysis (FEA) was performed using ABAQUS^TM to calculate the stress and deformation of a disk specimen. To determine the best mesh design for a thin disk specimen, hexahedron, hex-dominated, and tetrahedron models were used in FEA. All models revealed similar dome-shaped deformation behavior of the disk specimen. However, there was a considerable difference in stress distribution in the disk specimens. In the hex-dominated model, the applied stress was calculated to be the maximum at the dome center, whereas the stress was calculated to be the maximum at the dome edge in the hexahedron and tetrahedron models. From a comparison of the FEA results with deformation behavior and SCC location on the disk specimen after RDCT, the most proper FE model was found to be the tetrahedron model.

• 전기화학회지
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• 제10권1호
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• pp.25-30
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• 2007

고분자전해질 연료전지의 성능은 cell 온도, 전체 압력, 반응 기체의 부분 압력 상대습도와 같은 다양한 요인들에 의해 영향을 받는다. 이온화된 수소 이온은 $H_3O^+$의 형태로 membrane을 통과하여 물을 생성하는 반응으로 전기를 발생시킨다. 대용량 연료전지에서는 부수적으로 생성되는 열을 제거하거나 다른 용도로 사용할 목적으로 냉각시스템이 필요하다. 냉각수의 전도도가 상승할 경우에 연료전지에서 발생된 전류의 일부가 냉각수를 통하여 누설되어 연료전지의 성능을 감소시킬 수 있다. 본 연구에서는 3차 증류수와 ethylene glycol이 함유되어 있는 부동액을 사용하여 저항 수치 변화를 관찰하는 실험을 수행하였다. 3차 증류수의 경우 저항값이 설정치 이하로 내려가는데 약 28일이 소요되었고, 연료전지의 운전에 의한 영향은 관찰되지 않았다. 부동액을 냉각수로 사용한 경우는 43일이 지나도 저항값이 설정치 이하로 내려가지는 않았지만, stack 분리판의 접착부에 이상이 생긴 것으로 추정되는 연료전지의 성능 저하가 발생하여 전도도 실험을 중단하였다. 고분자전해질 연료전지에서는 수소이온의 이온전도성 저하를 방지하기 위하여 외부에서 가습하여 주는 방식이 일반적이지만, 소용량 연료전지에서는 무가습 조건을 적용하여 연료전지의 효율을 높이고 제작단가도 경감할 수 있다. 이를 위하여 저가습 및 무가습 실험을 수행하였으나 대용량 연료전지에서는 양측 무가습인 경우에 $50{\sim}60^{\circ}C$ 이상의 고온에서 성능이 발현되기 어려운 것으로 관찰되었다. 냉각수의 유량을 다르게 하여 실험을 수행한 경우에는 0.78L/min과 같은 낮은 유량에서 출구온도와 입구온도를 측정하여 본 결과 두 온도 사이에 ${\Delta}T$가 다른 유량에서보다 크게 발생하여 성능이 감소된 것으로 사료된다. 이와 같이 냉각수의 온도와 유량을 다르게 하여 양측 무가습 실험을 수행한 결과, 연료전지의 성능이 cell 온도에 직접적인 연관이 있는 것으로 관찰되었다.

• 한국안전학회지
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• 제20권2호
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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.