• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2010년도 제35회 추계학술대회논문집
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• pp.9-12
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• 2010

본 논문은 연료 과농 연소 환경 하의 이중 와류 동축형 분사기의 유량 통과 특성 파악을 위해 수행한 실험결과를 수록하였다. 액체산소와 케로신(Jet A-1)을 사용하여 연소시험을 수행하고 유량 통과 특성을 유량계수로 표현하였다. 유량계수 산출을 위해 유량, 압력, 온도를 계측하였다. 연료 분사기의 경우, 산화제 측 분사기 형상, 연소압, 혼합비에 관계없이 일정한 유량 계수 값을 보였다. 이에 반해 산화제 분사기는 연소압과 혼합비 변화에 영향을 받는 것으로 나타났다. 화염 형성 변화가 유량계수 변화에 특히 산화제 측에 영향을 주고 있음을 밝혔다.

• 한국연소학회:학술대회논문집
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• 한국연소학회 2014년도 제49회 KOSCO SYMPOSIUM 초록집
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• pp.179-182
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• 2014

A comparative DNS study of the ignition characteristics of dual-fueled reactivity controlled compression ignition (RCCI) and stratification charge compression ignition (SCCI) is investigated using a 116-species reduced primary reference fuel (PRF) mechanism. In the RCCI combustion, two PRF fuels (n-heptane and iso-octane) with opposite autoignition characteristics are separatedly supplied and in-cylinder blended such that spatial variations in fuel reactivity, fuel concentration and temperature are achieved. In the SCCI combustion, however, just a single fuel (PRF50) is used such that only fuel concentration and temperature inhomoginieties are obtained. Because three factors, rather than only two as in SCCI combustion, govern the overall RCCI combustion, combustion timing and combustion duration or heat release rate of RCCI combustion are flexibly and effectively controlled. It is found that the overall RCCI combustion occurs much earlier and its combustion duration is longer compared to SCC combustionI. Moreover, the negative temperature coefficient (NTC) has a positive effect on enhancing RCCI combustion by inducing a shorter combustion timing and a longer combustion duration as a result of the occurrence of a predominant low-speed deflagration-combustion mode.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• 한국자동차공학회논문집
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• 제21권1호
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• pp.23-29
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• 2013

In recent years, development of energy conversion systems using hydrogen as an energy source has been accelerated globally. Even though hydrogen is an environment-friendly energy source, safety and effectiveness issues in storage, transportation, and usage of hydrogen should be clearly resolved in every application. Therefore, sensors for detecting hydrogen leakage, especially for fuel cell electric vehicles, should be designed to have much higher resolution and accuracy in comparison with conventional gas sensors. In this study, we conducted to determine the design parameters for the semiconductor hydrogen sensor with optimized sensing conditions under the thermal distribution characteristic and thermal transfer characteristic. The heat generation study on power supply voltage was studied for correlation analysis of thermal energy according to the power supply voltage variation from 1.0 voltage to 10.0 voltage every 0.5 voltage. And we studied for the temperature coefficient of resistance with hydrogen sensor.

• Nuclear Engineering and Technology
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• 제20권3호
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• pp.189-196
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• 1988

핵연료소결체의 편심이 정상상태에서 핵연료봉 열적 성능에 미치는 영향을 조사하였다. 지배방정식은 핵연료소결체와 피복관영역에 대해 2차원 원통좌표계 (r, $\theta$)로 각각 세우고 유한요소법으로 풀었다. 갭(gap)영역에서 방위각 의존적인 열전달계수를 사용하여 동심구조는 그대로 두는 반면 갭크기의 비대칭성을 고려하였다. 재료물성치는 온도의 함수로 사용되었으며 체적 열발생은 반경의 함수로 고려하였다. 핵연료 소결체의 편심으로 인해 피복관 외부 표면에서 최대국부열속은 증가하였고, 핵연료 소결체의 최대온도와 핵연료 평균온도는 감소하였다. 전자는 최소 DNBR계산시 불확실도에 영향을 미칠 것으로 생각되며, 후자의 두현상은 핵연료 소결체의 용융 가능성과 사고시 핵연료 잠재에너지를 줄어들게 할 것으로 예상된다. 또한, 핵연료 소결체의 편심으로 인해 핵연료 소결체의 온도분포는 비대칭을 이루고 최대온도의 위치는 변동되었다.

• 대한기계학회논문집
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• 제10권1호
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• pp.7-14
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• 1986

본 연구에서는 가압경수형 원자로심을 모의하는 3*3배열로 된 모의연료 봉다발의 실험장치를 이용하여 재완수과정의 유동특성과 열전달특성을 파악하였으며, 재완수과정중 연료봉의 온도거동을 예측하는 REFLUX코드를 최근 개발된 연구자료를 토대로 수정하여 본실험결과와 비교하였다.

• 한국자동차공학회논문집
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• 제10권4호
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• pp.15-22
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• 2002

As the emission regulations on the automobiles have been increasingly stringent, precise control of air/fuel ration is one of the most important issues on the gasoline engines. Although many researches have been carried out to identify the fuel transport phenomena in the port fuel injection gasolines, mixture preparation in the cylinder has not been fully understood due to the complexity of fuel film behavior, In this paper, the mixture preparation during cold engine start is studied by using a Fast Response Flame ionization Detector.(FRFID) In order to estimate the transportation of injected fuel from the intake port into cylinder, the wall wetting fuel model was used. The two coefficient($\alpha$,$\beta$) of the wall-wetting fuel model was determined from the measured fuel mass that was inducted into the cylinder at the first cycle after injection cut-in. $\alpha$( ratio of directly inducted fuel mass into cylinder from injected fuel mass) and $\beta$ (ratio of indirectly inducted fuel mass into cylinder from wall wetted fuel film on the wall) was increased with increasing cooling water temperature. To reduce a air/fuel ratio fluctuation during cold engine start, the appropriate fuel injection rate was obtained from the wall wetting fuel model. Result of air/fuel ratio control, air/fuel excursion was reduced.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2048-2054
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• 2022

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.