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In the numerical simulation of a fluid at supercritical pressure flowing upward in vertical heated channels, difficulties arise due to the dramatic variation of physical properties, especially density, when fluid temperature crosses the pseudocritical temperature. Many different turbulence models were tried to simulate the fluid flo w under strong property (especially density) variation, but never succeeded in reproducing experimental data. This failure is considered to be caused by ineptness of the presently-available turbulence models in the simulation of strong property-varying fluids. In the present study a new eddy viscosity model is presented, which adopted a different approach in using the length and time scales rather than the velocity and length scales. The new eddy viscosity model also incorporated the influence of density gradient. With the new models, fluid flows accompanying strong density gradient were successfully simulated, and satisfactorily agreed with the experimental data.
Influeza type A virus have been worldwide problematic in animals as well as in humans. In this study, the use of reverse-transcriptase polymerase chain reaction (RT-PCR) was described for detecting influenza virus type A. The primer of RT-PCR was designed from an nonstructural (NS) gene of Influenza A virus. By RT-PCR, a product with the size of 189 bp was detected only when influenza virus type A was used as template. No products could be detected with Influenza virus type B as well as other respiratory pathogens. The detection limit of the RT-PCR was up to 100.3TCID50 which is comparable to the sensitivity of cell culture method. The RT-PCR could detect the influenza A virus from nasal turbinates of the ferrets infected with influenza virus type A not type B.
Korea Atomic Energy Research Institute (KAERI) has conducted steady-state tests with the constant heat flux conditions to investigate the scale effect on heat removal behavior of the air-cooled Reactor Cavity Cooling System (RCCS). The heated cavity height of the test facility at KAERI is 4.0 m, a quarter-scale of the full scale cavity height. The post-test system-analysis results using GAMMA+ was performed to simulate air-cooled RCCS test results . GAMMA+ with improved heat transfer model showed the good predictability for the test results. The governing mechanisms are the radiation across the cavity and the convection in the riser duct, based on the comparison with test ana system-analysis results. Especially, the mixed convection in the riser duct is very important to extrapolate RCCS coolability from test results. Additionally, the in-house computational fluid dynamics analysis code was used to find suitable turbulence models to simulate the KAERI test results.
Heat transfer experiments at a vertical annulus passage were carried out in the SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation) to investigate the heat transfer behaviors of supercritical CO₂. The collected test data are to be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). The mass flux was in the range of 400 ~1200 ㎏/㎡s and the heat flux was chosen up to 150 ㎾/㎡. The selected pressures were 7.75 and 8.12 ㎫. The heat transfer data were analyzed and compared with the previous tube test data. The test results showed that the heat transfer characteristics were similar to those of the tube in case of a normal heat transfer mode and degree of heat transfer deterioration became smaller than that in the tube. Comparison of the experimental heat transfer coefficients with the predicted ones by the existing correlations showed that there was not a distinct difference between the correlations.
SCWR (SuperCritical Water-cooled Reactor) is a feasible option for the 4th generation nuclear power plant. The main advantage of SCWR is very high thermal efficiency. A proper design requires a good description of heat transfer characteristics in relevant geometries and operating conditions. A recently built supercritical pressure test facility in KAERI has been used in producing experimental data on heat transfer and pressure drop in a flow of supercritical pressure CO₂ in geometries relevant to a proposed SCWR core. Currently, heat transfer experiment with a small diameter circular tube is being executed. The test result is introduced and compared with correlations published previously.
Safety Depressurization System of the Korean Next Generation Reactor prevents the Reactor Coolant System from over-pressurization by discharging the coolant with high pressure and temperature iota the In-containment Refueling Water Storage Tank(IRWST) during an accident. If temperature in the IRWST rises above the temperature limit of 200 of due to the discharged coolant, an unstable steam condensation may occur and cause large load on the IRWST wall. To investigate whether this condition can be reached or not for the design basis accident, the flow and temperature distributions of water in the IRWST were calculated by using CFX 4.2 computer code. The results show that the local water temperature does not exceeds the temperature limit within the transient time of 5 seconds.
한국형 차세대원자로에서는 비상노심 안전주입수가 저온관을 통하지 않고 원자 로용기에 직접 주입된다. 원자로용기의 가압열충격과 열수력적 관점에서 최적의 노즐위치를 결정하기 위해서 전산유체역학을 활용하였다. 상용 전산유체코드인 CFX를 이용하여 원자로 하향유로를 모사하는 해석대상 격자를 다중블록으로 형성한 다음 유동장을 비압축성 Navier-Stokes 운동량 방정식, 에너지 방정식과 표준 k-ε 난류모형 등으로 모형화하여 3차원 비정상상태 계산을 수행하였다. CFX에서는 경계 밀착좌표계 - 비엿물림격자와 SIMPLE 알고리즘을 사용한다. 본 연구결과 원자로용기의 가압열충격 관점에서 가장 보수적인 사고인 증기관 과단사고시에도 열적혼합이 잘 일어나 가압열충격이 발생할 가능성이 없는 것으로 판단되며 안전주입수 노즐이 저온관 바로 위에 위치할 때 원자로 하향유로 내의 온도 분포가 가장 균일하여 열적 혼합 관점에서는 최적의 위치로 판단된다.
Vortex type Fluidic Device(FD) which is installed at the bottom of Safety Injection Tank(SIT) controls the discharge flow rate from the tank. In case of loss of coolant accident the injection water flows into primary system in two steps; initial high flow rate for certain period of time and subsequent low flow rate. By two-step control of the discharge flow rate, FD can ensure the effective use of water in the tank. A small-scale FD has been tested to obtain a required flow characteristics maintaining full pressure and height of prototype, which are the major contributing parameters. Through the testing of many different arrangements of internal geometry of FD, most appropriate one was selected and its performance data was obtained. As characteristics of FD, time dependent Euler number, flow rate and pressure are presented and discussed. Also a method to predict the full size FD is presented.
원자로계통수 조건(300℃, 160㎏/㎠)에서 방사성 부식생성물 중 ^(60)Co를 제거하기 위한 원전계통수정화용 고온흡착제의 흡착메카니즘 및 제조방법을 알아 보았고, Fe-Ti-O, TiO₂, ZrO₂ 등의 흡착제를 공침법, 금속알콕사이드가수분해법, 졸겔법에 의해 제조하였다. 고온수에서 이들 흡착제의 Co^(2+) 흡착특성을 Autoclave를 이용한 회분식 흡착실험으로 살펴 보았으며, Fe-Ti-O 흡착제 제조시 소결온도 변화에 따른 Co^(2+) 흡착용량과 TiO₂ 흡착제 제조시 pH 변화에 따른 흡착용량 및 비표면적 등을 알아 보았다. To remove the solube corrosion products, mainly ^(60)Co under PWR reactor coolant conditions(300℃, 160㎏/㎠), the adsorption mechanism and preparation method of high-temperature adsorbents for reactor water purification were studied. Fe-Ti-O, TiO₂ and ZrO₂ adsorbents were prepared by coprecipitation, hydrolysis of metal alkoxide and sol-gel method. The Co^(2+) adsorption characteristic of these adsorbents in high temperature water were investigated in batch adsorption experiment by a stirred autoclave. The effect of sintering temperature on Co^(2+) adsorption capacity of Fe-Ti-O adsorbent and the effect of preparation pH on Co^(2+) adsorption capacity and specific surface area of TiO₂ adsorbent are reported.