• 한국환경과학회지
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• 제20권5호
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• pp.623-630
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• 2011

This study investigated the degradation of N, N-Dimethyl-4-nitrosoaniline (RNO, indicator of the generation of OH radical) by using dielectric barrier discharge (DBD) plasma. The DBD plasma reactor of this study consisted of a quartz dielectric tube, titanium discharge (inner) and ground (outer) electrode. The effect of shape (rod, spring and pipe) of ground electrode, diameter (9~30 mm) of ground electrode of spring shape and inside diameter (4~13 mm) of quartz tube, electrode diameter (1~4 mm), electrode materials (SUS, Ti, iron, Cu and W), height difference of discharge and ground electrode (1~15.5 cm) and gas flow rate (1~7 L/min) were evaluated. The experimental results showed that shape of ground electrode and materials of ground and discharge electrode were not influenced the RNO degradation. The thinner the diameter of discharge and ground electrode, the higher RNO degradation rate observed. The effect of height gap of discharge between ground electrode on RNO degradation was not high within the experimented value. Among the experimented parameters, inside diameter of quartz tube and gas flow rate were most important parameters which are influenced the decomposition of RNO. Optimum inside diameter of quartz tube and gas flow rate were 7 mm and 4 L/min, respectively.

• 대한기계학회논문집B
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• 제30권6호
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• pp.497-504
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• 2006

A Coriolis mass flowmeter(CMF), which has U-Shaped unique measurins tube was developed fo. direct mass flow measurement. In order to convert the time difference between two measuring tubes motion into mass flowrate and flow quantity, a signal processing circuit, as a part of CMF, was also developed. The CMF was designed as the 15 mm nominal diameter of pipe connection and the 8 mm stainless steel(sus 316) pipe was used for measuring tube. To maximize the flow signal(time difference) from the measuring tubes, the natural frequency of measuring tube was adjusted as 220 Hz, which is same as the frequency of exciter. The maximum displacement at the end of the measuring tube was measured as 0.05 mm and the maximum time difference between two measuring tubes was observed as $20{\mu}s$, which was proper for discrimination and measuring range of CMF. The developed CMF was tested against the gravimetric flowmeter calibrator in the range of 3 kg/min and 30 kg/min. The results showed that the CMF has good linearity and repeatability in the tested flow range. Large size of CMF base on the current study experience will be developed.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.257-264
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• 2024

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2000년도 KIASC Conference 2000 / 2000년도 학술대회 논문집
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• pp.19-21
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• 2000

We are developing portable type HTS magnet system cooled by solid nitrogen. This system have recooling and recharging capabilities. In this paper, we report preliminary test results obtained from the experimental solid nitrogen system and pancake magnet would with Bi-2223/Ag tapes, respectively. The operation period was sensitively dependent on the vacuum rate n the cryostat, size of SUS tube for flowing N_{2}$, and liquid nitrogen to cool the cryostat. The fabricated coil I_{c}$was 75 A at 20 K in self field.

• 멤브레인
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• 제15권4호
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• pp.330-337
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• 2005

[ $250^{\circ}C$의 고온에서 수증기 선택 투과 특성을 가지는 silica 막을 메탄을 탈수에 의한 dimethyl ether (DME) 합성 반응에 분리막 반응기로 적용하였다. Silica 전구체로서 tetraethoxysilane (TEOS)을 이용하여 초음파 분무 열분해 및 기상화학 증착법(CVD)법 등에 의해 다공성 스테인레스 스틸(SUS)에 silica 막을 합성하였다. CVD법에 의해 합성한 silica막의 수증기 투과도 및 메탄올에 대한 분리계수 상관관계 trade-off 선이 열분해 silica 막보다 높이 존재하였다. 수증기 투과도가 $1.2\times10^{-7}\;mol\;{\cdot}\;m^{-2}\;{\cdot}\;S^{-1}\;{\cdot}\;Pa^{-1}$ 이상이고, 메탄올에 대한 분리계수가 10 이상의 성능을 가지는 분리막 반응기에 대해서 기존 반응기 대비 $20\%$ 이상 메탄을 전환율이 향상되었다. 고온 수증기 선택성 silica 막이 메탄을 탈수 반응에 의해 생성되는 수증기를 제거함으로서 촉매 활성 저하를 억제하여 반응 전환율을 개선시키는 막 반응기로서의 효과를 확인할 수 있었다.

• 대한기계학회논문집B
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• 제41권10호
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• pp.639-647
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• 2017

Solid SCR에 사용 가능한 암모니아 저장물질의 하나인 암모니움 카바메이트는 열 분해시 이산화탄소 가스와 암모니아 가스를 생성하며, 분해 온도인 $60^{\circ}C$ 이하에서 암모니움 염으로 재결합되는 단점이 있다. 이러한 재결합 현상을 극복하기 위하여, 희석기체인 압축공기를 이용하여 기초가시화 실험을 수행하였다. 또한, 재결합 현상을 계량화하기 위하여, 재결합 물질의 무게변화를 측정하기 위한 간단한 장치를 만들어 자동차환경에서 사용되는 SUS재질의 3가지 관경에 크기에 대한 상관관계를 검토하였다. 아크릴 튜브로 제작된 온도조절이 가능한 가시화 실험장치에, 암모니아 가스, 이산화탄소 가스, 희석기 체인 질소 가스를 공급하며, 재결합 방지를 위한 온도, 압력, 희석유량과의 관계를 고찰하고, Chapman-Enskog Theory에서 파생된 Diffusivity를 사용하여 재결합 조건을 유추할 수 있는 지표로 사용하고자 한다.

• Journal of Advanced Marine Engineering and Technology
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• 제32권5호
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• pp.724-729
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• 2008

Umbilical cord blood has been recently considered an attractive potential alternative as a source of stem cell transplantation to curing diseases such as leukemia, cancers, immune disorders. Normally the stored system of the umbilical cord blood specimen is equipped with a computer-controlled robotic arm that enables the samples to locate the identification places in liquid nitrogen tank at regulated temperature as about $-196^{\circ}C$. As the half of robotic arm is in the air and the rest part is submerged in liquid nitrogen, the temperature of robotic arm varies from ambient to liquid nitrogen temperature. In this study the temperature variation of upper part of arm above tank lid was thermally analysed by using the commercial code of Ansys. The result of analysis was that the upper part of robotic arm was seriously frozen due to heat transfer from liquid nitrogen as low as -$120^{\circ}C$. In order to solve the frost problem of robotic arm, small PTFE tube block as resistance material was introduced into the lower part of tank lid instead of the whole stainless steel(SUS) robotic arm. The results showed that the temperature of robotic arm above the lid was higher enough, and this method would be one of the very effective measure to solve the problem.

• 전기화학회지
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• 제7권2호
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• pp.94-99
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• 2004

연료극 지지체 평관형 고체산화물 연료전지(SOFC)의 셀 전력밀도를 증가시키기 위하여 압출법에 의하여 제조하고 그 특성을 연구하였다. 연료극 지지체로써 Ni/YSZ($8mol\%$ yttria stabilized zirconia) cermet는 기공율 $50.6\%,\;0.23{\mu}m$의 기공크기를 나타내었다. 지지체에서의 Ni의 분포는 균일하였으며 전자전도 경로로써의 Ni의 연결성은 양호하였다. 지지체에 YSZ전해질과 복합 공기극층인 $LSM((La_{0.85}Sr_{0.15})_{0.9}MnO_3)/YSZ$ 복합층, LSM, LSCF $(La_{0.6}Sr_{0.4}Co_{0.2}Fe_{0.8}O_3)$층이 슬러리 디핑법에 의하여 코팅 및 소결된 연료극 지지체식 평관형 고체산화물 연료전지 단위전지의 성능은 $800^{\circ}C$에서 $300mW/cm^2(0.6V,\;500mA/cm^2)$의 성능을 나타내었다. 임피던스 분석에 의하여 평관형 셀의 전기화학적 분극저항을 평가하고 연료측의 가습에 따라 분극저항이 감소되어 성능이 향상됨을 알 수 있었다 슬러리 디핑법으로 LSM이 코팅된 SUS430 금속연결재를 $Ar+10\%\;H_2$에서 소결하였으며, $750^{\circ}C$에서 면저항의 측정할 결과, 초기에는 $148m{\Omega}cm^2$를 나타내었으며, 450시간 경과 후에 $43m{\Omega}cm^2$의 낮은 면저항을 유지하였다. 반면에 동일한 조건으로 LSM이 코팅된 Fecralloy는 높은 면저항을 나타내었다.