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Effect of Molten Corium Behavior Uncertainty on the Severe Accident Progress
Choi, Wonjun,Kim, Taeseok,Jeon, Joongoo,Kim, Nam Kyung,Kim, Sung Joong Hindawi Limited 2018 Science and technology of nuclear installations Vol.2018 No.-
<P>Uncertainty of a severe accident code output needs to be handled reliably considering its use in safety regulation of nuclear industry. In particular, severe accident codes are utilized for probabilistic safety assessment (PSA), where the uncertainty of severe accident progress should be considered carefully due to its influence on human reliability analysis. Therefore, in this study, the uncertainty analysis of severe accident progress was performed using MELCOR code, and a total of 200 data sets of in-vessel uncertainty parameters were generated by Latin hypercube sampling method. The rank regression analysis was also performed to investigate the effect of uncertainty parameters on the severe accident progress. Sensitivity coefficients (SCs) in MELCOR such as molten clad drainage rate and zircaloy melt breakout temperature showed significant influence on relocation time and dryout time of lower plenum. However, the influence of uncertainty parameter diminished as the accident progressed.</P>
Jeon, Joongoo,Choi, Wonjun,Kim, Nam Kyung,Kim, Sung Joong Elsevier 2018 Annals of nuclear energy Vol.121 No.-
<P><B>Abstract</B></P> <P>A safety injection (SI) flow model predicting target depressurization was developed in the previous study. The model estimated the sum of the decay heat and oxidation heat using the core exit temperature increase rate and core water level decrease rate during the accident progression. However, in the old model only the heat transfer to the coolant was considered but the heat accumulation in the structures was not included in detail. To resolve this issue, therefore, a new mechanistic model was developed by considering heat sources accumulated in the core heat structures. The accuracy of the new model was validated through the prediction of core total heat using the MELCOR 1.8.6 code. It was confirmed that the new model resulted in a relatively small error less than 10% in almost all sections while the old model exhibited a large error exceeding 50% since the start of oxidation for postulated SBO severe accident scenario. Through the model validation, an improved SI flow map was developed to predict more accurate target depressurization of the reactor coolant system (RCS) needed for maintaining core coolability. This study suggests that new SI flow map can effectively assist operator’s execution related to the RCS depressurization and SI injection into the RCS implemented in the severe accident management guideline under various severe accident scenarios.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Developed a safety injection (SI) flow model to determine the core coolability of OPR1000. </LI> <LI> Predicted injection flow rate to restore the core water level. </LI> <LI> Added detailed heat sources accumulated in heat structures in the model. </LI> <LI> Predicted accurate target depressurization of reactor coolant system to maintain core coolability. </LI> </UL> </P>
Kim, Taeseok,Choi, Wonjun,Jeon, Joongoo,Kim, Nam Kyung,Jung, Hoichul,Kim, Sung Joong Hindawi Limited 2018 Science and technology of nuclear installations Vol.2018 No.-
<P>During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.</P>
Revisiting the Regression between Raw Outputs of Image Quality Metrics and Ground Truth Measurements
JUNG, Chanho,JOO, Sanghyun,NAM, Do-Won,KIM, Wonjun INSTITUTE OF ELECTRONICS, INFORMATION & 2016 IEICE TRANSACTIONS ON INFORMATION AND SYSTEMS E SE Vol.99e.d No.11
<P>In this paper, we aim to investigate the potential usefulness of machine learning in image quality assessment (IQA). Most previous studies have focused on designing effective image quality metrics (IQMs), and significant advances have been made in the development of IQMs over the last decade. Here, our goal is to improve prediction outcomes of 'any' given image quality metric. We call this the 'IQM's Outcome Improvement' problem, in order to distinguish the proposed approach from the existing IQA approaches. We propose a method that focuses on the underlying IQM and improves its prediction results by using machine learning techniques. Extensive experiments have been conducted on three different publicly available image databases. Particularly, through both 1) in-database and 2) cross-database validations, the generality and technological feasibility (in real-world applications) of our machine-learning-based algorithm have been evaluated. Our results demonstrate that the proposed framework improves prediction outcomes of various existing commonly used IQMs (e.g., MSE, PSNR, SSIM-based IQMs, etc.) in terms of not only prediction accuracy, but also prediction monotonicity.</P>
Bench Scale급 석탄가스화기 시스템내의 고체시료 특성
정봉진(Jung, Bongjin),이나연(Lee, Na-Yeon),이찬(Lee, Chan),남원준(Nam, Wonjun),김경훈(Kim, Kyoung-Hoon),윤용승(Yoon, Young-Seung) 한국신재생에너지학회 2011 한국신재생에너지학회 학술대회논문집 Vol.2011 No.05
석탄가스화 복합발전(IGCC) 시스템은 고온 고압으로 운전되는 가스화기에서 미분탄을 산소와 함께 가스화하여 주로 CO 및 H₂를 생성하고 이때 발생되는 먼지 및 황성분은 각각 집진기 및 탈황장치에서 제거되며, 석탄 회분은 고온에서 용융되어 슬래그의 형태로 배출되는 방식을 사용하고 있다. 본 연구에서는 석탄가스화 복합발전시스템 설계에 필요한 기본자료를 파악하기 위해서, 고온 고압의 운전조건에서 1일 3톤의 석탄을 처리할 수 있는 Bench Scale급 석탄가스화기를 이용하여 가스화에 사용된 원탄 및 가스화기 설비의 각 지점에서 샘플링한 고체 시료를 중심으로 열화학적 특성을 살펴보았다. 가스화 실험은 아역청탄 계열의 ABK 석탄을 대상으로 가스화기 내부의 온도와 압력을 1400{sim}1450?C, 7.5{sim}7.6Kg/cm²로 유지시키면서 실시하였다. 실험에 사용된 석탄 시료의 기본적인 물성치를 조사하기 위하여 표준방법에 따라 석탄의 공업분석, 원소분석, 발열량분석 등을 실시하였다. 석탄가스화기에서 배출된 슬래그와 대상 석탄 회분의 특성을 파악하기 위해서 XRF를 이용한 회분의 성분분석, Heating Microscope를 이용한 회분의 용융점 분석, XRD를 이용한 회분과 슬래그내의 화합물의 형태 및 결정구조 파악, SEM을 이용한 슬래그의 형상 등을 분석하였다. 또한 석탄가스화기 시스템을 구성하는 각 설비의 특성을 파악하기 위해서 관련 설비의 특정 지점에서 채취한 시료의 입도분석, 원소분석, 촤 회분 무게비, 슬래그중의 잔존탄소함량, 슬래그와 슬래그로부터 제조된 용출수내의 중금속 함량분석 등을 실시하였다.