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인공하수 조성 성분에 따른 SBR 처리 공정의 효율에 관한 연구
이장훈,장승철,권혁구,김동욱,Lee Jang-Hoon,Jang Seung-Cheol,Kwon Hyuk-Ku,Kim Dong Wook 한국환경보건학회 2005 한국환경보건학회지 Vol.31 No.2
The removals of organic matter, nitrogen and phosphate in wastewater were investigated with Sequencing Batch Reactor (SBR). Glucose and sodium acetate were Used for organic carbon source so as to know nutrient removal efficiency in proportion to MLSS concentration. In the case of glucose, the COD removal rate was $74\%,\;41\%\;and\;66\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was $57\%,\;21\%\;and\;38\%$, the T-N was $24\%,\;13\%\;and\;44\%$, and the T-P was $12\%,\;21\%\;and\;33\%$. As a result, the removal rate of organic materials showed the finest remove when MLSS was 5000, but the nutrient removal rate appeared as was best when MLSS was 1000. In the case of sodium acetate, the COD removal rate was $83\%,\;81\%\;and\;86\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was appeared by $76\%,\;82\%\;and\;92\%$, the T-N $57\%,\;42\%\;and\;78\%$, and the T-P $48\%,\;52\%\;and\;38\%$. As a result, organic and T-N removal rates were best when MLSS was 1000. But, the T-P removal rates were best when MLSS was 3000. Glucose was shown fast removal in reaction beginning, but screened by more efficient thing though sodium acetate removes organic matter, nitrogen and phosphate. Form of floc was ideal in all reactors regardless of carbon source and MLSS concentration. And its diameter was about $200\~500{\mu}m$.
국내 연구용원자로 전출력 내부사건 1단계 확률론적안전성평가
이윤환,장승철,Lee, Yoon-Hwan,Jang, Seung-Cheol 한국안전학회 2021 한국안전학회지 Vol.36 No.3
This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.
리스크정보 최적화를 통한 국내 연구용원자로의 안전성 향상
이윤환,장승철,Lee, Yoon-Hwan,Jang, Seung-Cheol 한국안전학회 2022 한국안전학회지 Vol.37 No.2
This paper describes an attempt to improve and optimize the operational safety level of a domestic research reactor by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety at an operating research reactor in Korea, to evaluate whether it is probabilistically safe and reliable to operate, and to obtain insights regarding the requisite procedural and design improvements for achieving safer operation. The technical objectives were to use the PSA to identify the accident sequences leading to core damage, and to conduct sensitivity analyses based thereon to derive insights regarding potential design and procedural improvements. Based on the dominant accident sequences identified by the PSA, eight types of sensitivity analysis were performed, and relevant insights for achieving safer operation were derived. When these insights were applied to the reactor design and operating procedure, the risk was found to be reduced by approximately ten times, and the safety was significantly improved. The results demonstrate that the PSA methodology is very effective for improving reactor safety in the full-power operating phase. In particular, it is a highly suitable approach for identifying the deficiencies of a reactor operating at full power, and for improving the reactor safety by overcoming those deficiencies.
국내 원자력발전소의 주제어실 화재피난 리스크 평가를 위한 화재 시뮬레이션
강대일 ( Dae Il Kang ),김길유 ( Kilyoo Kim ),장승철 ( Seung Cheol Jang ),유성연 ( Seong Yeon Yoo ) 한국안전학회(구 한국산업안전학회) 2014 한국안전학회지 Vol.29 No.4
In this paper, to systematically assess the abandonment risk of main control room (MCR) fire, fire simulations with FireDynamics Simulator were performed and abandonment probabilities were estimated for the MCR bench-board fire of domestic referencenuclear power plant. The fire simulation scenarios performed in this study included propagating and non-propagating fires of the MCRbench-board, and the availability and unavailability of heating, ventilation, and air conditioning system (HVACS). The following resultswere obtained. First, temperature was the major abandonment impact factor for the MCR bench-board fire if the HVACS was availableand optical density was that if the HVACS was unavailable. Second, the fire scenario contributing the MCR bench-board fireabandonment risk was identified to be only the propagating fire. Third, it was confirmed that the abandonment probability of the MCRbench-board fire for domestic reference nuclear power plant could be reduced by using the fire modeling.
김재환(Kim Jae Whan),정원대(Jung Wondea),장승철(Jang Seung Cheol),곽상록(Kwak Sang Log) 한국철도학회 2006 한국철도학회 학술발표대회논문집 Vol.- No.-
The railway human reliability analysis (R-HRA) plays a role of identifying and assessing human failure events in the framework of the probabilistic risk assessment (PRA) of the railway systems. This paper reviews three existing HRA methods including the K-HRA (THERP/ASEP-based) method, the HEART method, the RSSB-HRA method, and introduces a case study that was performed to select an appropriate method for a railway risk assessment. The case is the signal passed at danger (SPAD) events, which are caused from a variety of factors. From the case study, the strengths and limitations of each method were derived and compared with each other from the viewpoint of the applicability to the railway industry.
신화재 확률론적안전성평가 방법 적용: 정성적 분석 결과
강대일(Gang, Dae-Il),김길유(Kim, Gil-Yu),장승철(Jang, Seung-Cheol) 한국화재소방학회 2013 한국화재소방학회 학술대회 논문집 Vol.2013 No.춘계
이 논문에서는 신화재 확률론적 안전성평가 (PSA) 방법 중 정성적 분석 방법을 울진 3호기 원전에 적용한 결과를 기술하였다. 지금까지 대부분의 국내 원전 에서는 EPRI 화재 PSA 방법을 이용하여 화재 PSA를 수행해 왔었다. 최근 미국 규제기관과 산업체에서는 신화재 PSA 방법으로 NUREG/CR-6850을 개발하였다. 신화재 PSA 방법을 이용하여 울진 3호기를 정성적으로 분석한 결과 150개의 방화지역 중 75개 지역이 정량적 분석 대상으로 파악되었다. 이는 기존 EPRI 화재 PSA 방법으로 수행한 방화지역 수보다 23개 많았다. 또 화재 PSA 수행을 위한 기기 수는 770여개이고 케이블 수는 6,000여개로 나타났다.