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중수로 실증 실험설비를 이용한 소형냉각재상실사고의 MARS-KS 입력모델 개발 및 검증계산
백경록,유선오,Baek, Kyung Lok,Yu, Seon Oh 한국안전학회 2021 한국안전학회지 Vol.36 No.2
Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.
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OLGA를 이용한 해저 가스 유전의 Slugging 예측
백경록(Kyung-Lok Baek),박상민(Sang-Min Park),박진후(Jin-Hoo Park),장광필(Kwang-Pil Chang) 대한조선학회 2011 대한조선학회 학술대회자료집 Vol.2011 No.6
As a recognized critical issue in design and operation of subsea oil/gas systems, flaw assurance is an engineering analysis process to assure that hydrocarbon fluids are transferred economically and safely from resemoir to the end user during the life of a project. One of the flow assurance challenges is prediction of the slugging in the subsea pipelines which could potentially block the flow path of hydrocarbon fluids because slugging greatly affects the design of receiving facilities such as separators and slug catchers. Tire facilities can be flooded and damaged if the slug is larger than the receiving capacity of the facilities. Tlw'efore, quantifying the slug volume is a essential part in the design of subsea production plants. The pui'pose of this paper is to analyze the effect of Operating conditions on slug volume in the subsea pipeline system using multiphase flaw simulation code, OLGA. It is shewn that the slug volumes into the separator through the subsea pipelines are affected by the operating conditions, ramp-up scenarios, and tend to be created lower for the conditins of applying the two step ramp-up scenarios rather than the single step and linear ramp-up scenarios.
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중수로 원전 가상의 mSGTR과 SBO 다중 사건에 대한 MARS-KS 코드 분석
유선오(Seon Oh YU),이경원(Kyung Won LEE),백경록(Kyung Lok BAEK),김만웅(Manwoong KIM) 한국압력기기공학회 2021 한국압력기기공학회 논문집 Vol.17 No.1
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.
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