• Nuclear Engineering and Technology
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• 제45권4호
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• Journal of Mechanical Science and Technology
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• 제19권9호
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• pp.1781-1789
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• 2005

In order to realistically predict the combustion characteristics of the oxy-fuel flame, the present study employs the non-adiabatic flame let approach. In this combustion model, the detailed equilibrium chemistry is utilized to accurately account for the thermal dissociation as well as to properly include the radiative cooling effects on the detailed chemistry. Numerical results indicate that the present approach has the capability to correctly capture the essential features and precise structure of the oxy-fuel flames. In this work, the detailed discussion has been made for the characteristics of oxy-fuel flames, the capability and defect of the present approach and also uncertainties of experimental data.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.509-522
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• 2011

FRAPCON-3.4a and FRAPTRAN 1.4 are the most recent versions of the U.S. Nuclear Regulatory Commission (NRC) steady-state and transient fuel performance codes, respectively. These codes have been assessed against separate effects data and integral assessment data and have been determined to provide a best estimate calculation of fuel performance. Recent updates included in FRAPCON-3.4a include updated material properties models, models for new fuel and cladding types, cladding finite element analysis capability, and capability to perform uncertainty analyses and calculate upper tolerance limits for important outputs. Recent updates included in FRAPTRAN 1.4 include: material properties models that are consistent with FRAPCON-3.4a, cladding failure models that are applicable for loss-of coolant-accident and reactivity initiated accident modeling, and updated heat transfer models. This paper briefly describes these code updates and data assessments, highlighting the particularly important improvements and data assessments. This paper also discusses areas of improvements that will be addressed in upcoming code versions.

• 대한기계학회논문집A
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• 제34권9호
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• pp.1217-1225
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• 2010

소듐을 냉각재로 사용하는 고속로의 증기발생기에서는 소듐과 물의 화학적 반응을 최소화하는 것이 중요한 문제이다. 소듐과 물의 반응 가능성을 줄여 증기발생기의 신뢰성을 향상시키기 위한 한가지 방안으로 이중벽관을 전열관으로 사용하는 증기발생기를 개발하고 있다. 이 증기발생기에서 중요한 현안은 이중벽관에서의 열전달 성능을 향상시키는 문제와 원자로 운전 중에 소듐과 물 반응사고가 일어나기 전에 전열관의 파손을 감지하는 기술을 개발하는 것이다. 이 논문에서는 이 현안을 극복할 수 있는 방안을 제시하였고, 이 기술을 활용하여 증기발생기의 개념을 설계하였다. 또한 이 개념에 적용되는 이중벽관을 설계 및 예비 제작하여 기계적 시험을 수행하였다.

• 대한기계학회논문집
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• 제16권4호
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• pp.787-794
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• 1992

본 연구에서는 이미 알려진 이상적인 단열모델에 대한 해석적 근사해를 기본 해로 취하고, 열교환과정의 손실 및 작동유체의 유동손실 등 성능에 미치는 영향이 비교적 큰 인자만을 고려하여, 성능을 쉽게 예측할 수 있는 2차단열해석법의 하나를 개발하고자 한다. 방법의 타당성및 적용예를 보이기 위하여 기존의 스터링기관중 각 종제원및 실험결과가 발표되어 있어 비교의 기준으로서 적합한 GPU-3(ground power unit)기관을 대상으로 해석방법을 실제 적용하고 결과를 고찰하기로 한다.

• Nuclear Engineering and Technology
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• 제31권5호
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• pp.465-475
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• 1999

The objective of the present work is to assess the analysis capability of two wall film condensation models, the default and the alternative models, of RELAP5/MOD3.2 on condensation experiments in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. In the calculation of a base case the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, and Its alternative model over-predicts them throughout the condensing tube, Also, both models over-predict the void fractions. The nodalization study shows that the variation of the node number does not change both modeling results of RELAP5/MOD3.2 Sensitivity study for varying input parameters shows that the inlet steam-air mixture flow rate, the inlet air mass fraction, and the inlet saturated steam temperature give significant changes of their heat transfer coefficients Run statistics show that the grind time of the default model is always higher than that of the alternative model by about 23%.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3379-3397
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• 2021

In this study, a multi-physics modeling method was developed to analyze a nuclear fuel rod's thermo-mechanical behavior especially for high temperature anisotropic creep deformation during ballooning and burst occurring in Loss of Coolant Accident (LOCA). Based on transient heat transfer and nonlinear mechanical analysis, the present work newly incorporated the nuclear fuel rod's special characteristics which include gap heat transfer, temperature and burnup dependent material properties, and especially for high temperature creep with material anisotropy. The proposed method was tested through various benchmark analyses and showed good agreements with analytical solutions. From the validation study with a cladding burst experiment which postulates the LOCA scenario, it was shown that the present development could predict the ballooning and burst behaviors accurately and showed the capability to predict anisotropic creep behavior during the LOCA. Moreover, in order to verify the anisotropic creep methodology proposed in this study, the comparison between modeling and experiment was made with isotropic material assumption. It was found that the present methodology with anisotropic creep could predict ballooning and burst more accurately and showed more realistic behavior of the cladding.

• 한국추진공학회지
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• 제26권6호
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• pp.10-20
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• 2022

본 연구에서는 케로신 다단연소사이클 엔진용 예연소기를 설계하기 위해, 고압의 산화제 과잉 조건에서 예연소가스를 계산하고 냉각유로에서 극저온 유체의 복합열전달 및 수력 특성을 해석할 수 있는 설계코드를 개발하였다. 사용자 편의성과 범용성을 가진 오픈 소스 라이브러리 Cantera를 활용하였으며, 실제유체의 열역학/전달 상태량을 정확히 계산하기 위해 관련 소스 코드들을 새로 작성하여 Cantera에 추가하였다. 현재 예비설계 중인 100톤급 부스터 엔진용 예연소기에 적용하였으며, CFD 해석결과와 비교를 통해 설계코드로서의 예측 정확도와 활용성을 확인하였다.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2004년도 학술대회 논문집
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• pp.375-380
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• 2004

This paper presents the basic quench protection idea for the HTS(High-Temperature Superconducting) cable. In Korea power system, the transfer capability of transmission line is limited by the voltage stability, and HTS cable could be one of the countermeasure to solve the transfer limit as its higher current capacity and lower impedance[1]. However, the quench characteristic of HTS cable makes HTS cable to loss its superconductivity, and therefore change the impedance of the line and power system operating condition dramatically. This pheonominum threats not only HTS cable safety but also power system security, therefore a proper protection scheme and security control counterplan have to be established before HTS cable implementation. In this paper, the quench characteristics of HTS cable for the fault current based on heat balance equation was established and a proper protection method by FCL(Fault Current Limiter) was suggested.