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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.
RD-14M LOCA 실증실험 검증계산을 위한 MARS-KS 해석 모델 개발
유선오(Seon Oh Yu) 대한기계학회 2021 大韓機械學會論文集B Vol.45 No.10
원전 안전해석에 사용되는 전산코드의 예측성능은 주요 과도/사고를 모의하는 실증실험의 검증계산을 통해 평가된다. 본 연구에서는 MARS-KS 코드의 중수로 사고해석 적용 유효성 확인을 위한 코드 검증평가 기반 마련을 위해 RD-14M의 LOCA 실증실험(B9006)을 모의하는 해석 모델을 개발하였고, 실험 결과와 비교·분석하였다. 코드는 주요 실험 변수의 전반적인 열수력 거동을 유사하게 모의하였고, 계통 내 물리적 현상을 합리적으로 예측하는 것으로 평가되었다. 그리고 전산코드의 사고해석 적용 유효성 확인을 위해 코드 version 간 실증실험 예측성능 변화를 상호 비교하였다. 본 논문은 RD-14M 실증실험의 검증계산을 위한 해석 모델과 MARS-KS 코드 성능의 개선에 기여할 것으로 기대된다. The prediction performances of computational codes for safety analyses were confirmed by validation of the tests simulating the key transients/accidents of nuclear power plants. In this study, an analysis model of the RD-14M LOCA test (B9006) was developed to evaluate code validation and confirm the availability of the MARS-KS code for accident analyses of CANDU-6 plants. It was observed that the code appropriately simulated the overall thermal-hydraulic behaviors and physical phenomena of the major parameters. Moreover, to confirm the code applicability to the CANDU-6 accident analyses, the prediction variations in experiments between code versions were compared. This study is expected to contribute toward improving the analysis model and code performance for validation calculation of the RD-14M test.
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.
Turbulence Induced Fretting Wear Characteristics of Steam Generator Helical Tubes
정명조(Myung Jo Jhung),조종철(Jong Chull Jo),김효정(Hho Jung Kim),윤영길(Young Gill Yune),유선오(Seon Oh Yu) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.5
This study addresses safety assessment of the potential for fretting wear damages on steam generator helical tubes due to turbulence induced vibration in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for helical type tubes with various conditions. Special emphases are put on the effects of coil diameter and the number of turns on the modal and fretting wear characteristics of tubes. Also, investigated are the effects of external pressure on the tube modal characteristics as well as the effects of turbulence induced vibration on the fretting wear characteristics of tubes.