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Analysis of Dose Rates from Stream Generators to be Replaced from Kori Unit 1
Son,Jung-Kwon,Shin,Sang-Woon,Cho,Chan-Hee,Song,Myung-Jae 대한방사선 방어학회 1998 방사선방어학회지 Vol.23 No.3
1998년에 고리 1호기로부터 교체될 증기발생기의 선량율을 계산하기 위하여 Smear 오염검사결과와 튜브 손상원인을 규명하기 위하여 인출하였던 증기발생기 튜브의 선량율 측정결과로부터 증기발생기 내부의 방사성핵종 재고량을 평가하였다. 방사성 핵종 재고량을 토대로 QAD-CG 컴퓨터 코드를 이용하여 증기발생기 표면의 접촉 선량율과 1m 이격 선량율을 계산하였는데, 접촉 선량율은 Channel Head의 하부에서 최저인 11.5 mR/hr를 나타냈으며, shell barrel의 중간 지점에서 최대값인 37.7 mR/hr를 나타냈다. 한편 접촉 선량율과 1 m 이격 선량율은 증기발생기의 크기로 인해 큰 차이를 보이지 않았다. 또한 증기발생기의 차폐가 필요한 경우 요구되는 기본적인 데이터를 마련하기 위하여 납과 탄소강의 차폐 특성을 비교해 보았다. 납을 차폐체로 사용할 경우 2 mm 두께만으로도 증기발생기 shell barrel 중간 지점에서의 표면 선량율이 37.7 mR/hr에서 15.7 mR/hr로 감소되었다. 그러나 탄소강의 경우에는 차폐체의 두께를 2 cm로 증가시킨다고 하더라도 차폐효과가 매우 낮았다. 이러한 차폐효과 차이는 저에너지 광자에 대한 납과 탄소강의 감쇄 효과 차이와 축적인자 차이 때문에 발생되는 것으로 추정된다. In order to calculate dose rates from steam generators to be replaced from Kori unit 1 in 1998, radionuclide inventories inside steam generator were evaluated from smear test results and measured dose rates from S/G tubes withdrawn for the metallographical examination of damaged tubes. Based on the inventories, contact dos rates and dose rates at 1 m from the surface of a steam generator were calculated using the QAD-CG computer code. Contact dose rates ranged from 11.5 mR/hr at the bottom of channel head to 37.7 mR/hr at the middle of shell barrel, and showed no significant difference with dose rates at 1 m from the surface of steam generator. Shielding effects of lead and carbon steel were compared to provide basic shielding data. Lead shield showed excellent shielding effects. Dose rate at 1 m from the middle of S/G shell barrel decreased from 38.6 mR/hr to 15.5 mR/hr with the lead shield of 2 mm thickness. However, carbon steel showed a poor shielding effect even with the thickness of 2.0 cm. This can be explained with the great differences in the attenuation effect and buildup factor between lead and carbon steel for low energy photons.
이준(J. Lee),강한옥(H. O. Kang),서재광(J. K. Seo),유승엽(S. Y. Ryu),윤주현(J. Yoon),김긍구(K. K. KIM) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.11
In this paper, the thermal hydraulic design characteristics of the power conversion system for SMART 330 MWt plant which the once-through helically-coiled steam generators have been installed was analyzed. The feedwater temp., total feedwater flow rate, steam pressure, total steam flow rate and steam temp. was analyzed vs. steam generator power on conditions of a constant feedwater temp. and steam pressure, and was compared with the commercial nuclear power plants. After the present analysis, it was evaluated that at the range of low power(about below 20%) the steam temp. is nearly same as the core outlet temp., and at the range of meddle power(about between 20 ~ 80%) it show gradually a upward trend, and at the range of high power(about between 80 ~ 100%) it show gradually a downward trend. These phenomena result from the once-through steam generator's model which consists of the 3 different type of regions, i.e., subcooled, boiling, and superheated region.
김우곤,김종민,김민철 한국압력기기공학회 2020 한국압력기기공학회 논문집 Vol.16 No.1
Creep rupture data for Alloy 690 steam generator tubes in a pressurized water reactor are essentially needed to demonstrate a severe accident scenario on thermally-induced tube failures caused by hot gases in a damaged reactor core. The rupture data were obtained using the tube specimens under different applied-stress levels at 650˚C, 700˚C, 750˚C, 800˚C, and 850˚C. Important creep constants were proposed using various creep laws in terms of Norton power law, Monkman-Grant (M-G) relation, damage tolerance factor (λ), and Zener-Hollomon parameter (Z). In addition, a creep activation energy (Q) value for Alloy 690 tube was reasonably determined using experimental data. Creep behaviors such as creep strength, creep rates, rupture elongation showed the results of temperature dependence well. Modified M-G plot improved a correlation of the creep rate and rupture life. Damage tolerance factor for Alloy 690 tubes was found to be λ =2.20 in an average value. Creep activation energy for Alloy 690 tube was optimized for Q=350 (kJ/mol). A plot of Z parameter obeyed a good linearity, and the same creep mechanism was inferred to be operative in the present test conditions.
인코넬 합금의 미세조직과 기계적 특성에 미치는 냉각속도 영향
박노경,이호성,채영석,Park, No-Kyeong,Lee, Ho-Seong,Chai, Young-Suck 한국재료학회 2007 한국재료학회지 Vol.17 No.10
The mechanical properties and microstructure of Inconel 690 and 600 alloys with various cooling rates were investigated. Optical microscopy and scanning electron microscopy observations indicated that in case of the cooling rate of $0.5^{\circ}C/min$, discontinuous carbides along the grain boundaries were formed and when the cooling rate was $10^{\circ}C/min$, continuous carbides were formed in Inconel 690 and 600 alloys. For the annealed Inconel 690 alloy with high Cr content, a lot of annealing twins, which led the preferential growth of (111) planes, were observed. However, the annealed Inconel 600 alloy with low Cr content showed a few annealing twins and the preferential growth of (200) planes. Inconel 600 alloy had a larger value of ultimate tensile strength (UTS) than Inconel 690 alloy.
홍상범(Sang-Beom Hong),최준호(Jun-Ho Choi) 대한전기학회 2019 전기학회논문지 Vol.68 No.1
Combined Cycle Unit (CC) generates the primary power from the Gas Turbine(GT)and supplies the remaining heat of the GT to the Steam Turbine (ST) to generate the secondary power from the ST. It plays a major role in terms of energy efficiency and Load Frequency Control(LFC). Incremental Heat Rate (IHR) curves of economic dispatch(ED) of CC is applied differently by GT/ST combination. But It is practically difficult because of performance test by all combinations. This paper suggests a reasonable method for estimating IHR curves for partial combinations(1 : 1∼ (N - 1) : 1) using IHR curves when operating with GT alone(1:0) and with all (N : 1) combinations of CC .
김형진(Kim, Hyung-Jin),성봉주(Sung, Bong-Zoo),박치용(Park, Chi-Yong),유기완(Ryu, Ki-Whan) 한국소음진동공학회 2006 한국소음진동공학회 논문집 Vol.16 No.9
A modified energy method for the fretting wear of the steam generator tube is proposed to calculate the wear-out depth between the nuclear steam generator tube and its support. Estimation of fretting-wear damage typically requires a non-linear dynamic analysis with the information of the gap velocity and the flow density around the tube. This analysis is very complex and time consuming. The basic concept of the energy method is that the volume wear rate due to the fretting-wear phenomena Is related to work rate which is time rate of the product of normal contact force and sliding distance. The wearing motion is due to dynamic interaction between vibrating tube and its support structure, such as tube support plate and anti-vibration bar. It can be assumed that the absorbed work rate would come from turbulent flow energy around the vibrating tube. This study also numerically obtains the wear-out depth with various wear topologies. A new dissection method is applied to the multi-span tubes to represent the vibrational mode. It turns out that both the secondary side density and the normal gap velocity are important parameters for the fretting-wear phenomena of the steam generator tube.