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CFD를 이용한 소듐냉각고속로 열교환기 설계코드의 검증
김대희(D. Kim),어재혁(J.H. Eoh),윤정(J. Yoon),이태호(T.H. Lee) 한국전산유체공학회 2013 한국전산유체공학회 학술대회논문집 Vol.2013 No.5
Korea Atomic Energy Research Institute is developing a SFR(Sodium-cooled Fast Reactor) which is one of the generation IV nuclear reactors. The SFR is composed of four main fluid systems which are categorized by its functions, i.e, PHTS(Primary Heat Transport System), IHTS(Intermediate Heat Transport System), SGS(Steam Generation System) and DHRS(Decay Heat Removal System). The coolant of the SFR is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, PHTS to IHTS and PHTS to DHRS. For the design and performance analysis of the heat exchangers, a computer code have been written. In this work, the computer code named SHXSA is validated through the CFD simulations.
전산유체역학을 이용한 소듐 소듐 열교환기 설계코드의 검증
김대희(D. Kim),어재혁(J.H. Eoh),이태호(T.H. Lee) 한국전산유체공학회 2016 한국전산유체공학회지 Vol.21 No.1
A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the 4th generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named SHXSA has been validated preliminarily by computational fluid dynamics simulations.
액체금속로 증기발생기에서 대규모 누출 소듐-물 반응 사고시 중후반기 압력거동특성 연구
정지영,어재혁,정경채,김병호,김태준,남호윤,박남국 한국공업화학회 2005 응용화학 Vol.9 No.1
In order to investigate the pressure transient at the later phase of a large leak sodium-water reaction event in KALIMER(Korea Advanced Llquid MEtal Reactor), an experimental study has been carried out. The experiment was performed to simulate the long-term system transient response of a large leak sodium-water reaction event due to double ended guillotine failure of three tubes in a 1/10 scale-down mock-up test facility with sufficient considerations of the quasi-steady state features of the SWR event.