• 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.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.407-408
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• 2002

This paper presents a split cooling system for a new inline 4-cylinder automotive engine. The split cooling system circulates coolant to the cylinder head and cylinder block separately. The coolant flow in the cylinder block is controlled by a $2^{nd}$ Thermostat installed at the outlet of cylinder block. The $2^{nd}$ thermostat closes when the coolant temperature is low. And this makes the coolant flow in cylinder block nearly stagnant, thereby reducing the coolant-side heat transfer coefficient and raising cylinder bore temperature. The $2^{nd}$ thermostat starts to open when the coolant temperature reaches a specified temperature. The test results on engine dynamometer show improved fuel economy and lower exhaust emission which result from the decrease in friction works and cooling loss. Also, several vehicle tests, with application of the new engine have been performed. Fuel economy improvement of 0.5{\sim}2.0%$ yields from different test modes and details are discussed in this paper.

• 한국자동차공학회논문집
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• 제22권7호
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• pp.57-62
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• 2014

This study presented the feasibility of a coolant heat-source heat pump system as an alternative heating system for electrically driven vehicles. Heat pumps are among the most environmentally friendly and efficient heating technologies in residential buildings. In various countries, electric mobiles devices such as EV, PHEV, and FCEV, have been mainly concerned with heat pumps for new mobile markets. The experiments herein were conducted for various ambient temperatures and coolant temperatures to reflect the winter season. The system, a coolant heat-source heat pump, consisted of an inside heat exchanger, an outside heat exchanger, a motor driven compressor, an electronic expansion valve, and plumbing parts. For the experimental results, the maximum heating capacity and air discharge temperature are up to 6.3 kW and $62^{\circ}C$ respectively at an ambient temperature of $10^{\circ}C$, and coolant at $10^{\circ}C$. However, at $-20^{\circ}C$ ambient temperature and $-10^{\circ}C$ coolant temperature, conditions were insufficient to warm the cabin as the air discharge temperature was $13^{\circ}C$.

• 설비공학논문집
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• 제11권6호
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• pp.835-845
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• 1999

The objective of the present study was to investigate heat and mass transfer characteristics in a vertical falling film type absorber using LiBr-$H_2O$ solution with 6owt%. The experimental apparatus consisted of an absorber with inner diameter of 17.2 mm and length of 1150mm, a generator, an evaporator/condenser, a solution tank, a sampling trap etc. The parameters were solution temperature of 45 and $50^{\circ}C$, coolant temperature of 30 and $35^{\circ}C$, and film Reynolds numbers from 50 to 150. Pressure drop in the absorber increased as solution and coolant temperatures decreased. Pressure drop in the absorber increased up to the film Reynolds number of 90, and then decreased at the further increase of the Reynolds number above 90. The maximum absorption mass flux observed at the film Reynolds number of 90. Absorption mass flukes increased as coolant temperature decreased. Absorption mass fluxes and heat transfer coefficients under subcooled condition were larger than those under superheated condition. Heat transfer coefficients were affected by solution temperature more than coolant temperature. The maximum absorption effectiveness under the subcooled condition was 23% for coolant temperature of $30^{\circ}C$ and 31% for coolant temperature of $35^{\circ}C$ under the present experimental conditions.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.811-824
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• 1995

CANDU 핵연료봉의 휨 열적 휨 멘트와 수력학적 견인력 및 기계적 하중에 기인하는 휨 모멘트에 의하여 일어난다. 여기서, 연료봉 휨은 연료봉 축방향 중심선으로부터의 측면 처짐으로 정의한다. 본 논문에서는 연료봉 축방향 중심선에 대한 비대칭 온도불포에 의해 핵연료 피복관 자체와 피복관과 소결체의 상호작용 부위에서 발생하는 열적 휨만을 취급한다. 이를 위해 1).소결체와 피복관사이의 기계적 상호작용을 무시한 조건에서의 핵연료 피복관의 휨과 2) 소결체와 피복관의 온도 변화에 기인하여 발생하는 소결체와 피복관 사이의 기계적 상호작용을 고려한 조건에서의 연료봉 휨을 혼합 고려하고, 각각에서 피복관의 비대칭 온도분포가 (i) 냉각재의 불완전한 혼합에 따른 비균질 냉각재 온도, (ii) 핵연료 피복관과 냉각재 사이의 비균질한 열전달 계수, (iii) 핵연료내 반경 방향으로의 중성자속 감쇄에 의한 비대칭 열 발생 등의 복합적효과에 의해 발생되는 것으로 고려하여 피복관의 대칭온도 분포까지 포함 할 수 있는 열적 휨의 일반적 해석 공식을 제시하였다. 본 휨 공식에 사용되는 모든 변수에 대한 민감도 분석을 통해, 핵연료봉 길이, 피복관 내경, 냉각재 평균 온도 및 변화 인자, 소결체 -피복관 기계적 상호 작용 인자, 중성자속 감쇄 인자, 핵연료 열팽창 계수, 피복관-냉각재 열전도 계수 등의 변화가 피복관 두께, 피복관-냉각재 열전달 계수, 피복관 열팽창 계수, 핵연료-피복관 열전달 계수 등의 변화보다 핵연료봉의 열적 휨에 상대적으로 더욱 영향을 미치는 것으로 밝혀졌다.

• Nuclear Engineering and Technology
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• 제12권1호
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• pp.9-18
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• 1980

냉각재 상실사고에 따르는 rewetting현상을 연구하기 위하여 대기압에서 단일가열관을 사용한 재관수 실험을 수행하였다. Yamanouchi 이론을 바탕으로한 1차원 및 2차원 열전도 해석을 본 실험조건과 일치시키기 위해 수정하여 실험결과와 비교 검토하였다. 하부재관수 해석에서는 unrewetted 구역에서 증기의 열전달이 고려 되어야 한다는 것을 알았다. 실험을 통해 revetting속도는 시험관의 초기벽온도, 냉각재 유량, 냉각재 온도에 따라 달라진다는 것을 알았다. rewetting 온도와 열전달 계수를 평가하기 위한 보다 나은 방법의 개발이 필요하다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• 한국추진공학회지
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• 제10권3호
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• pp.90-98
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• 2006

재생냉각 시스템의 냉각제로 사용하는 탄화수소계열 연료인 Jet A-1의 냉각특성과 강제대류 열전달 특성을 파악하기 위해 가열된 직선 튜브를 제작하였다. 냉각제의 입, 출구온도와 튜브의 외벽 온도를 측정하여 강제대류 열전달 특성을 고찰하였다. 결과를 통하여 Jet A-1을 사용한 경우에는 일정 유동조건에 대하여 burnout이 발생하는 최초의 벽 온도를 결정함으로써 냉각제가 수용할 수 있는 최대 열유속을 확인하였다. 냉각제의 유속과 입구온도가 대류열전달 특성에 직접적인 영향을 주는 인자임을 확인하였다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.289-290
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• 2002

Stainless steel ball bearings are used in the control element drive mechanism and driving mechanisms such as step motor and gear boxes for the integral nuclear reactor, SMART. The bearings operate in pressurized pure water (primary coolant) at high temperature and should be lubricated with only this water because it is impossible to supply greases or any additional lubricant since the whole nuclear rector system should be perfectly sealed and the coolant cannot contain ingredients for bearing lubrication. Temperature of water changes from room temperature to about 120 degree Celsius and pressure rises up to 15MPa in the nuclear reactor. It can be anticipated that the frictional characteristics of the ball bearings changes according to the operating conditions, however little data are available in the literature. It is found that friction coefficient of 440C stainless steel itself does not change sharply according to temperature variation from the former research, and the friction coefficient is about 0.45 at low speed range. In this research frictional characteristics of the assembled ball bearings are investigated. A special tribometer is used to simulate the axial loading and the bearing operating conditions, temperature and pressure in the driving mechanism in the nuclear reactor. Highly purified water is used as lubricant ‘ and the water is heated up to 120 degree Celsius and pressurized to 15MPa. Friction force is monitored by the torque transducer.