• 대한기계학회논문집A
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• 제37권10호
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• pp.1251-1259
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• 2013

본 연구에서는 소듐냉각 고속로 붕괴열교환기(DHX)의 고온 설계 및 크리프-피로 손상 평가를 수행하였다. 제 4 세대 소듐냉각 고속로의 능동 및 피동 잔열제거계통에 설치되는 DHX와 한국원자력연구원의 STELLA-1 시험루프에 설치된 DHX에 대해 상세설계 및 3D 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH와 RCC-MR 코드를 따라 크리프-피로 손상평가를 수행하였다. 크리프-피로 손상평가 결과에 기초하여 두 설계기준에 대해 비교 분석하고, 설계 기술기준의 보수성 이슈에 대해 토의하였다.

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.

• 대한기계학회논문집B
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• 제34권11호
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• pp.991-998
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• 2010

소듐냉각 고속로 내부기기 배치 변경에 의한 초기냉각 성능변화를 검토하기 위하여 중간열교환기의 수직배치가 다른 3개의 원자로를 대상으로 COMMIX-1AR/P 코드를 활용한 다차원 해석을 수행하였다. 원통좌표계의 중심축을 기준으로 원주방향의 1/4 부분만을 모델링하고 정상상태 및 과도상태 분석을 수행하여 IHX 수직배치 변화가 초기 냉각 특성에 미치는 영향을 분석하였고, DHX를 통한 후기 냉각 모드 개시 시점에 미치는 영향도 분석하였다. 분석 결과 IHX 수직배치 상승은 원자로 풀내부 자연 순환 유량을 증가시켜 초기 냉각과정에서 노심 최고 온도의 급격한 상승을 방지할 수 있으며, 초기냉각 성능을 향상시키기 위한 관성회전차의 가용설계재원의 범위도 확대시킨다. 또한 IHX 수직배치 상승은 후기냉각모드에 큰 영향을 주지 않으면서 초기냉각성능의 향상에 기여할 수 있을 것으로 사료된다.

• Journal of Microbiology and Biotechnology
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• 제24권5호
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• pp.714-718
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• 2014

Apoptosis is the process of programmed cell death executed by specific proteases, the caspases, which mediate the cleavage of various vital proteins. Elucidating the consequences of this endoproteolytic cleavage is crucial to understanding cell death and other related biological processes. Although a number of possible roles for caspase-6 have been proposed, the identities and functions of proteins that interact with caspase-6 remain uncertain. In this study, we established a cell line expressing tandem affinity purification (TAP)-tagged caspase- 6 and then used LC-MS/MS proteomic analysis to analyze the caspase-6 interactome. Eight candidate caspase-6-interacting proteins were identified. Of these, five proteins (hnRNP-M, DHX38, ASPP2, MTA2, and UACA) were subsequently examined by co-immunoprecipitation for interactions with caspase-6. Thus, we identified two novel members of the caspase-6 interactome: hnRNP-M and MTA2.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.78-84
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• 2016

The FHX (Forced-draft sodium-to-air Heat Exchanger) employed in the ADHRS (active decay heat removal system) is a shell-and-tube type counter-current flow heat exchanger with M-shape finned-tube arrangement. Liquid sodium flows inside the heat transfer tubes and atmospheric air flows over the finned tubes. The unit is placed in the upper region of the reactor building and has function of dumping the system heat load into the final heat sink, i.e., the atmosphere. Heat is transmitted from the primary cold sodium pool into the ADHRS sodium loop via DHX (decay heat exchanger), and a direct heat exchange occurs between the tube-side sodium and the shell-side air through the FHX tube wall. This paper describes the DHRS and the structural design of the FHX.

• 한국전산유체공학회지
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• 제21권1호
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ 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.

• 대한기계학회논문집B
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• 제33권6호
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• pp.418-426
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• 2009

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.