[en] The development of thermal hydraulic analysis code for Gas-Cooled Reactors (GCRs) and for annular fuel and its application to various types of nuclear reactors, and the assessment of the Korea Advanced Institute of Science and Technology (KAIST) Departure from Nucleate Boiling (DNB)-type theoretical Critical Heat Flux (CHF) model for rod bundles with non-uniform axial power shapes were investigated. Thermal hydraulic characteristics of thorium-based fuel assemblies with annular seed pins were analyzed using Thermal-Hydraulic analysis code for Annular Fuel (THAF) combined with Multichannel Analyzer for steady states and Transients in Rod Arrays (MATRA), and compared with those of existing thorium-based assemblies. This study investigates the possibilities of using annular fuel pins in a pressurized water reactor with emphasis on coolant flow distribution and heat transfer fraction in internal and external sub-channels. MATRA and THAF showed good agreements for the pressure drops at the internal sub-channels. Mass fluxes were high in inner sub-channels of the seed pins due to the grid form losses in the outer sub-channels. About 43% of heat generated from the seed pin flowed into the inner sub-channel. The remaining heat flowed into the outer sub-channel. The inner to outer wall heat flux ratio was approximately 1.2. Maximum temperatures of annular seed pins were slightly above 500 .deg. C. Minimum DNB Ratios (MDNBRs) of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Temperatures and enthalpies were higher in the inner sub-channels due to the fact that inter-channel mixing cannot occur in the inner sub-channels. A thermal-hydraulic analysis code for annular fuel-based Liquid Metal Reactors (LMRs) has been developed. About 41% of the heat generated from the fuel pin flowed into the inner sub-channel and the rest into the outer sub-channel. The inner to outer wall heat flux ratio was equal to approximately 1.44. A new 37 annular fuel design when operated at 50% higher power density resulted in about 1400 .deg. C lower hot spot fuel temperature with 26.5% higher mass flux in the inner sub-channel. GCRs have been highlighted as a promising option for next generation reactor technology. A thermal hydraulic analysis code for GCRs has been developed with a heat transfer model of a block element which is solved implicitly with the helium energy equation. Validation was carried out through comparison with both experimental and analytical results. A computation module for annular fuel rods has been coupled to the code for comparative analyses of an annular fuel-based block element. At normal operation, the annular fuel shows 80 .deg. C lower peak temperature than the solid fuel for the same power in Japan's High Temperature Engineering Test Reactor (HTTR), even though the pressure drop is higher in the annular fuel. A general CHF prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis at the conceptual design stage for a new Pressurized Water Reactor (PWR). In this study, the KAIST liquid sub-layer dryout CHF prediction model for DNB region has been implemented in a sub-channel analysis code, and investigated for the method's possible use in a rod bundle environment with various non-uniform axial power shapes. The KAIST model showed comparable prediction capability to Lin's method for bottom-, center-, and top-peaked heat flux shapes. The KAIST model, without any correction factors or empirical constants, turned out to be suitable to fulfill the needs for a basis of a general CHF prediction method as compared to Lin's method and Westinghouse-3 (W-3) correlation

[en] Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using MATRA_A combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and MATRA_A showed good agreements for the pressure drops at the internal subchannels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel, which implies the inner to outer wall heat flux ratio was approximately 1.2. When the relative annular seed pin power is higher, the inner to outer wall heat flux ratio and inner subchannel mass flux are lower. The maximum temperatures of the annular seed pins were slightly above 500 .deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels

[en] The predictions are in good agreements with the final safety analysis report, which implies the containment integrity is maintained during or after an accident like loss of coolant accident. In this study, the CONTEMPT-LT/028 was used to calculate the pressure and temperature, and in the follow-up study, CONTAIN 2.0 will be used for the pressure and temperature predictions in APR1400 reactors. Shin-Hanul Units 1 and 2 may possess different characteristics of peak pressure and temperature in containment following a large break loss-of-coolant-accident. To assess the important performance independently and to compare with prediction results presented in the final safety analysis report (FSAR) of Shin-Hanul Units 1 and 2 might be helpful to regulatory review for identifying validity of the FSAR. The end of blowdown (EOB) time during a LOCA could largely affect the peak pressure and temperature in the containment. This paper provides CONTEMPT-LT/028 prediction of the peak pressure and temperature of Shin-Hanul Units 1 and 2 following a large break loss-of-coolant-accident and compares with licensee's prediction results.

[en] Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal subchannels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 .deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels

[en] A thermal-hydraulic analysis code which is capable of modeling both internally and externally cooled annular fuel pins was developed. The coolant flow distribution in the annular fuel-based assemblies is adjusted by a pressure drop model allowing for conditions such as non-equal velocity and non-saturated phases. The heat transfer fraction is determined by the ratio of cross-sectional areas distinguished by the radius at which the first derivative of the temperature within the annular fuel equals zero. The code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. The heat transfer fraction difference between the code and RELAP was about 3.9%, and the Departure from Nucleate Boiling Ratio (DNBR) prediction of the code agreed well with the MIT's result in the region below 3 m. For the application of the code, thermal-hydraulics of thorium-based fuel assemblies loaded with annular seed pins were compared with those of the existing thorium-based assemblies. The pressure drop in the assembly generally increased in the case of annular fuel due to the larger wetted perimeter. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel. The minimum DNBRs (MDNBRs) of the annular fuel-based assemblies were higher than those of the existing ones. Because interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels

[en] Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using MATRA_A combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and MATRA_A showed good agreements for the pressure drops at the internal subchannels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 qC. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels

[en] More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

[en] A licensing application of Shin-Hanul Units 1&2 was submitted and is currently under KINS’ review. These units are with 1400 MWe capacity each and similar to Shin-Kori Units 3&4. The application includes radiological consequences following postulated accidents such as LOCA, SGTR, etc. In general, the most severe accident in nuclear power plants is regarded as LBLOCA when it comes to radiological impact. This paper provides RADTRAD prediction of the radiological consequences of Shin-Hanul Units 1&2 following a large break loss-of-coolant-accident and compares with licensee’s prediction results. Simulation calculations using RADTRAD can be useful for predicting radiological consequences after a loss of coolant accident in APR1400 reactors. The predictions are in good agreements with the final safety analysis report. In this study, the RADTRAD was used to calculate radiological consequences in the event of LOCA, and in the follow-up study, some non-LOCA will be analyzed with RADTRAD for the radiological consequences in APR1400 reactors.

[en] In conventional air-conditioning systems, when the circulated air from the air-conditioned space pass through the cooling coil in the air-conditioning system, the air is over-cooled to eliminate the moisture as well as to decrease the temperature. The purpose of this study is to test and evaluate performance of the cooling/reheating system which can save both cooling energy and reheating energy by exchanging heat between the cooled air and reheated air. Experimental apparatus consists of fan, ducts, polymer exchangers, cooling coil, electronic auxiliary heater and data acquisition system. Two types of polymer exchanger, plate type and dimple type, made of polypropylene for cooling/reheating system are designed. Heat transfer and dehumidification characteristics of system are tested. The results show that the energy saving is up to 40% in the range of present experimented conditions, and it decreases with increasing velocity, inlet temperature and specific humidity

[en] Clear signs of increasing interest in the nuclear power option in national energy mix are being observed worldwide. Construction of Olkiluoto 3 is underway in Finland and France announced the construction of Flamanville 3 and its policy of developing Gen-4 reactors and replacing current operating Gen-3 reactors by 2050. In US, energy Policy Act of 2005 gives rise to expectation of new construction and Bush administration initiated Global Nuclear Energy Partnership (GNEP) with the aim of expanded use of nuclear energy worldwide. UK has expressed the intention of maintaining nuclear option in its national energy policy through several policy papers. Continuous nuclear program is promoted in Japan, Korea and Russia. China and India have ambitious plan of constructing large amount of NPPs. The expanded use of nuclear energy in the future, however, should require appropriate plans and much effort for ensuring safety and security with both global and national contexts. This paper reviews future challenges to nuclear safety and security and presents some perspectives for ensuring safe and secure use of nuclear energy