• Nuclear Engineering and Technology
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• v.54 no.6
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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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• v.53 no.6
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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.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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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.

• Transactions of the Korean Society of Automotive Engineers
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• v.22 no.7
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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 $13^{\circ}C$.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.11 no.6
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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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• v.27 no.6
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• pp.811-824
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• 1995

The CANDU element bowing is attributed to actions of both the thermally induced bending moments and the bending moment due to hydraulic drag and mechanical loads, where the bowing is defined as the lateral deflection of an element from the axial centerline. This paper consider only the thermally-induced bending moments which are generated both within the sheath and the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element The generalized and explicit analytical formula for the thermally-induced bending is presented in con-sideration of 1) bending of an empty tube treated by neglecting the fuel/sheath mechanical interaction and 2) fuel/sheath interaction due to the pellet and sheath temperature variations, where in each case the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. As the results of the sensitivity calculations of the element bowing with the variations of the parameters in the formula, it is found that the element bowing is greatly affected relatively with the variations or changes of element length, sheath inside diameter, average coolant temperature and its variation factor, pellet/sheath mechanical interaction factor, neutron flux depression factor, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient and sheath and pellet thermal conductivities.

• Nuclear Engineering and Technology
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• v.12 no.1
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• pp.9-18
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• 1980

Reflood experiments under atmospheric pressure have been conducted with a single heated tube to investigate basically the rewetting phenomena following a LOCA. Experimental conditions are 180cm length of test tube, wall temperature range of 300-80$0^{\circ}C$, coolant flooding rate of 5-30cm/sec. and subcooling of 35-85$^{\circ}C$. Experiments show that the rewetting velocity is dependent on the initial wall temperature of test tube, coolant flow rate and coolant subcooling. It is required to develop the proper method to evaluate the rewetting temperature and the heat transfer coefficient.

• Nuclear Engineering and Technology
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• v.56 no.3
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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.

• Journal of the Korean Society of Propulsion Engineers
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• v.10 no.3
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• pp.90-98
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• 2006

The Heated Tube Facility(HIF) was fabricated to identify the forced convective heat transfer and the cooling characteristic for the hydrocarbon fuel(Jet A-1), which is used for the coolant of the regenerative cooling system. The forced convective heat transfer coefficient was calculated from the measured coolant and tube surface temperature. In case of using the Jet A-1, the maximum heat flux which the coolant can absorb was identified by determining the critical wall temperature generating the burnout on the fixed flow condition. The inlet bulk-temperature of the coolant has a direct influence on the forced convective heat transfer characteristic.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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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.