• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.605-610
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• 1997

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.941-942
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• 2011

A thermal equivalent circuit of IPMSM considering eddy current loss of PM and core loss of rotor is proposed. This thermal equivalent model is represented by the thermal resistances and thermal capacitances. In order to determine the factor of each parameter, a heating test is processed. Additionally, the eddy current loss of PM is calculated by a transient 3D finite element analysis. Finally, this thermal equivalent model is verified by a temperature test in a 25kW 12-pole/18-slot IPMSM with varying load.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.169-179
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• 2000

A detailed numerical analysis of the evolution of thermal stratification in a curved piping system in a nuclear power plant is performed. A finite volume based thermal-hydraulic computer code has been developed employing a body-fitted, non-orthogonal curvilinear coordinate for this purpose. The cell-centered, non-staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is prevented by the application of momentum interpolation method. The SIMPLE algorithm is employed for the pressure and velocity coupling, and the convection terms are approximated by a higher-order bounded scheme. The thermal-hydraulic computer code developed in the present study has been applied to the analysis of thermal stratification in a curved duct and some of the predicted results are compared with the available experimental data. It is shown that the predicted results agree fairly well with the experimental measurements and the transient formation of thermal stratification in a curved duct is also well predicted.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2002년도 학술대회 논문집
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• pp.50-55
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• 2002

The transient numerical analysis was performed for vapor cooled current leads. The present numerical modeling considered that there is temperature difference between the copper lead and the helium vapor flow. This numerical modeling was compensated and validated by the experiment with commercially available 100 A current leads. The numerical modeling in this paper described thermal characteristics of overloaded current leads more accurately than the conventional steady state analysis. Proper design of overloaded current leads was suggested by indicating the appropriate overloading factor in the pulse mode operation.

• 전산구조공학
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• 제8권2호
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• pp.141-145
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• 1995

이 논문은, 선형점탄성체가 과도온도상태에 있을때의 거동을 시간영역 경계요소법에 의해 해석하고 있다. 점탄성체에 대해서 '열유동단순'거동을 가정하였다. 경계요소 공식화과정을 설명한 후, 예제문제에 대한 수치해석 결과를 보여주었다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.1974-1987
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• 2023

The load-following operation of small modular reactors (SMRs) requires accurate prediction of transient behaviors that can occur in the balance of plants (BOP) and the nuclear steam supply system (NSSS). However, 1-D thermal-hydraulics analysis codes developed for safety and performance analysis have conventionally excluded the BOP from the simulation by assuming ideal boundary conditions for the main steam and feed water (MS/FW) systems, i.e., an open loop. In this study, we introduced a lumped model of BOP fluid system and coupled it with NSSS without any ideal boundary conditions, i.e., in a closed loop. Various methods for coupling boundary conditions at MS/FW were tested to validate their combination in terms of minimizing numerical instability, which mainly arises from the coupled boundaries. The method exhibiting the best performance was selected and applied to a transient simulation of an integrated NSSS and BOP system of a SMART. For a transient event with core power change of 100-20-100%, the simulation exhibited numerical stability throughout the system without any significant perturbation of thermal-hydraulic parameters. Thus, the introduced boundary-condition coupling method and BOP fluid system model can expectedly be employed for the transient simulation and performance analysis of SMRs requiring daily load-following operations.

• 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.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.9-16
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• 1986

1984년 11월 14일 원자력 1호기에서 발생된 주급수 상실사고에 대한 계통의 열수력학적인 거동을 모의·해석하고, 발전소 실측자료와의 비교를 통하여 사용된 전산코드의 신뢰도를 평가하였다. 모의된 열수력학적 변수들은 발전소 실측자료와 비교적 잘 일치하였으나 원자로 트립시에 증기발생기 증기유량과 주 냉각재 계통 평균온도에 있어서 약간의 차이를 보였다. 이는 원자로 트립시 깎은 시간에 급격한 노심 출력의 감소로 인하여 열·수력학적 변수들에 큰 변화를 야기하여 발전소 실측자료가 과도상태에서의 불학실성을 내포하기 때문으로 예측되었다. 해석에 사용된 전산코드는 RELAP5/MOD1/CY018로부터 불합리한 oscillation을 일으키는 interphase drag 및 wall heat transfer model의 수정을 통하여 개발된 RELAP5/MOD1/NSC이다.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.40-49
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• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1376-1381
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• 2004

Understanding the thermal conductivity and heat transfer processes in superlattice structures is critical for the development of thermoelectric materials and optoelectronic devices based on quantum structures. $Chen^{(1)}$ developed ballistic diffusive equation(BDE) for alternatives of the Boltzmann equation that can be applied to the complex geometrical situation. In this study, a simulation code based on BDE is developed and applied to the 1-dimensional transient heat conduction across a thin film and transient 2-dimensional heat conduction across the film with heater. The obtained results are compared to the results of the $Chen^{(1)}$ and Yang and $Chen^{(1)}$. Finally, steady 2-dimensional heat conduction in the quantum dot superlattice are solved to obtain the equivalent thermal conductivity of the lattice and also compared with the experimental data from $Borca-Tasciuc^{(2)}$.