• 식물분류학회지
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• 제35권1호
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• pp.47-55
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• 2005

백두산 및 만주 일대에 생육하는 17과 26속 29종류 31개 집단의 관속식물에 대하여 어린 꽃봉오리를 채취하고 이를 고정하여 감수분열상을 관찰하여 이들 지역에 생육하는 관속식물의 세포분류학적 특징을 논의하고자 하였다. 백두산 지역에서 채집된 종류는 두메양귀비, 등대시호, 땃두릅나무, 가는다리장구채, 비로용담 등 28 집단에 속하는 26종류이었으며 두만강변에서 채집된 종류는 현삼, 과꽃, 신감채 등 3종류로 나타났다. 본 연구를 통하여 비단쑥은 2n=9II로, Cacalia komaroviana는 2n=30II로, 바위구절초는 2n=27II로, 그리고 좁은잎돌꽃은 2n=11II로 그 염색체의 수가 새롭게 보고되었으며, 개발나물과 현삼의 경우는 새로운 염색체의 수가 보고되었다. 기타 기존의 보고와 염색체 수가 일치하는 종류들 중 도깨비엉겅퀴, 나비나물, Allium strictum 그리고 짚신나물의 경우는 4배체로 나타났으며, 두메양귀비는 6배체로 나타났다. 아울러 오랑캐장구채, 털부처꽃 및 물매화의 경우 2배체로 나타났다.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.12-19
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• 1990

1981년 6월 9일 고리 1호기 원자력발전소에서 발생한 외부 전원 상실사고 자료를 근거로 RELAP5/MOD2코드모델 평가를 하였다. 계산된 주요 열ㆍ수력학 변수를 실측자료와 비교 분석하였으며 증기발생기의 Nodalization 민감도 분석이 수행되었다. 계산된 열ㆍ수력학 변수는 실측치와 비교적 잘 일치하고 있으며, 이러한 유형의 사고 분석에 RELAP5/MOD2가 적합하다는 것을 보였다. 그러나 가압기 압력과 수위변동에서는 상당한 차이를 보였으며 높게 계산되었다. 이러한 사실은 RELAP5의 수직관에서의 층류 열전달 모델에 기인하는 것으로 해당모델의 개선을 요하고 있다는 것을 알았다. 그리고 증기발생기의 Nodalization 연구를 통하여 수위변동을 잘 예측하기 1위해서는 증기발생기 증기 Dome와 Downcomer사이에 압력을 전달시켜주는 유로를 모델링 하여야 한다는 것을 알았다.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• 식물분류학회지
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• 제33권4호
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• pp.455-472
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• 2003

수우(樹友) 이창복교수는 1946년 서울대학교에 부임하여 1984년 퇴임하기까지 우리 나라 식물분류학계의 기반을 다지는데 큰 기여를 하였다. 특히, 1952년부터 1984년까지 33년간 남한지역에서 약 7만여 점의 표본을 채집하여 서울대학교 농업생명과학대학 수우(樹友)표본관(SNUA)의 근간을 확립하여 체계적인 국내 최대의 표본관으로 발전시켰다. 시기별 주요 채집 활동으로 1950년대는 소지 식별을 위한 채집과 박사학위 논문연구를 위한 채집, 1960년대는 1920-1930년에 걸쳐 일본학자인 T. Nakai가 채집한 희귀식물에 대한 분포 확인차원의 채집을 시도한 시기였고, 1970년대와 1984년까지 학술조사 목적으로 특정지역의 식물상을 조사하는 채집이 주를 이루었다. 이창복교수는 참나무속과 싸리속 표본에 대한 채집과 기타 목본, 희귀식물과 관련된 수집을 직접 현장에서 식물을 확인하고 확증 표본 확보에 주력하여 현재 SNUA의 소장표본이 전국 주요 지역 식물상의 대표성을 가지는데 많은 공헌을 하였다. 특히 이창복교수의 1950년 후반에서 1960년 초반에 걸쳐 채집하여 연구한 아시아산 낙엽성 참나무속에 대한 결과는 많은 관련 연구에 근간이 되고 있다. 그러나, 불행히도 표본을 제외한 기타 채집기록이 별로 남아 있지 않아 주요 채집지와 채집 날짜 이외에 조사와 채집의 목적을 명확하게 검증하기 어렵다.

• Journal of Radiation Protection and Research
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• 제32권2호
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• pp.79-88
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• 2007

원자력발전소 사고 시 방사성물질이 발전소외 지역으로 유출이 되거나 예상될 때 주민을 안전하게 보호하기 위한 조치의 일환으로 주민소개가 고려된다. 소개시간 산정에 필요한 인자를 도출하고, 각각의 인자에 대해 원전 주변의 현장자료를 토대로 부지주변의 교통환경 여건을 반영하여 월성원전의 방사선 비상계획구역 내의 주민전체를 소개시키는데 소요되는 시간을 예측하였다. 월성원전 방사선비상계획구역 내 주요 간선도로와 교차로에서 교통량을 조사하였고, 소개시작시간 분포를 추정하기 위해 상주거주자와 일시거주자를 대상으로 사회행동특성에 대한 설문조사를 실시하였다. 평시와 관광객이 많이 유입되어 차량정체가 예상되는 여름철을 대상으로 주간 및 야간, 평상기상 및 악기상의 경우로 나누어 주민소개시간을 예측하였다. 주민 소개시간 예측을 위한 교통분석은 TSIS 패키지 프로그램이 이용되었다. 비상계획구역 경계 남단과 북단에서 모든 소개차량(인구)이 비상계획구역을 벗어나는 데 걸리는 시간은 전체적으로는 $118{\sim}150$분 정도로 예측되었다. 여름 첨두교통량 유입시, 소개 시간은 낮이 밤보다 최대 17분 정도 더 소요되는 것으로 예측되었다.

• 한국안전학회지
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• 제29권6호
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• pp.166-171
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• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• 전기학회논문지P
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• 제65권4호
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• pp.235-241
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• 2016

Grid-connected, large-capacity energy storage systems (ESS) can be used for peak load supply, frequency regulation, and renewable energy output smoothing. In order to confirm the capability of battery ESS to provide such services, 4MW/ 8MWh battery ESS demonstration facility was built in the Jocheon substation on Jeju Island. The frequency regulation technology developed for the Jocheon demonstration facility then became the basis for the 28MW and 24MW frequency regulation ESS facilities installed in 2014 at the Seo-Anseong and Shin-Yongin substations, respectively. The operation control systems at these two facilities were continuously improved, and their successful commercialization led to the construction of additional ESS facilities all over Korea in 2015. In seven (7) locations nationwide (e.g., Shin-Gimje and Shin-Gyeryeong), a total of 184 MW of ESS had been commercialized in 2016. The trial run for the new ESS facilities had been completed between April and May in 2016. In this paper, results of the trial run from each of the ESS facilities are presented. The results obtained from the Seo-Anseong and Shin-Yongin substations during a transient event by a nuclear power plant trip are also presented in this paper. The results show that the frequency regulation battery ESS facilities were able to quickly respond to the transient event and trial run of ESS is necessary before it is commercialized.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.232-235
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• 2008

The vibration of steam turbine is caused by Mass Unbalance, Shaft Misalignment, Oil Whip and Rubbing etc. but in turbine which is normally operated and maintained, the Mass Unbalance component possesses the greatest portion. Our power plant has two steam turbines in capacity of 200MW and 135MW respectively and each turbine is supported by 6 journal bearings. However, we had many difficulties because the vibration amplitude of No 3 and 4 Bearings was high during the start-up and operation mode change of steam turbine. But, with this study, we completely solved the vibration problem caused by the mass unbalance of No 1 steam turbine. Until a recent date, No 3 and 4 bearings which support high pressure turbine for No 1 steam turbine had shown about 135${\mu}$m in vibration amplitude (sometimes it increased to 221${\mu}$m maximum. alarm: 6mils, trip: 9mils) at base load. After applying the study, they decreased to about 40${\mu}$m maximum. It is a result from that we did not change the setting value of Bearing Alignment and only changed the assembly position of internal parts in Synchro Clutch Coupling Rachet Wheel which links between high pressure turbine and low pressure turbine, and increased the internal gap and machining of the Pawl stopper surface. In the operation of steam turbine, if the vibration value increases by 1X, we should reduce the vibration of bearing by weight balancing. However, unless the vibration of bearing is declined by the balancing, we will have to disassemble and check the component and find the cause. In this study, We researched the way to lower mass unbalance that is 1X vibration component which has the greatest portion of vibration generated by steam turbine and We got good result by applying the findings of this study.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.