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성대현 ( D. H. Seong ),이광원 ( K. W. Rhie ),김태훈 ( T. H. Kim ),오동석 ( D. S. Oh ),오영달 ( Y. D. Oh ),서두현 ( D. H. Seo ),김영규 ( Y. G. Kim ),김은정 ( E. J. Kim ) 한국안전학회(구-한국산업안전학회) 2012 한국안전학회지 Vol.27 No.3
This study is about the quantitative safety assessment of hydrogen station in Korea operating with on-site type. This was written by background information that before qualitative safety assessment to write. For the qualitative safety assessment method, the study used FMEA(failure mode & effect analysis) and HAZOP(hazard & operability), and adopted the FTA(fault tree analysis) as the quantitative safety assessment method. To write the FTA, we wrote FT by Top event that hydrogen leakage can be called most serious accident of hydrogen station. Each base event collect reliability data by reliability data handbook, THERP-HRA and estimation of the engineering. Assessment looked at the high frequency and the possible risk through Gate, Importance, m.cutsets analysis
원자력수소생산용 공정열교환기 개발을 위한 고온/고압 소형가스루프설계
홍성덕(S. D. Hong),오동석(D. S. Oh),김종호(J. H. Kim),김용완(Y. W. Kim),이원재(W. J. Lee),장종화(J. H. Chang) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.6
We designed a small scale gas loop that can simulate reference operating conditions, that is, temperature up to 950 ℃ and pressure up to 6㎫. Main objective of the loop is to screen the candidate process-heat-exchanger designs of very small capacity of 10~20㎾. We arranged the components of primary gas loop and secondary SO₃ loop. Design requirements are prepared for the safe design of a main heater, a hot-gas-duct and a process heat exchanger that is excluded a risk of a failure owing to the thermal stresses, the flow-induced-vibration or the acoustic vibration in both nitrogen and helium mediums. We determined a total pressure loss of the primary loop to 83 ㎪ and the minimum outer diameter of the loop pressure pipe to 90㎜ at hot location that will guarantee from a thermal failure. Very toxic SO₃ secondary loop is needed a scrubber and a SO₃ collector for safety and preventing a contamination of environment.
CFD 방법을 이용한 핵연료다발내의 복합 유동혼합 날개의 최적화
인왕기(W. K. In),오동석(D. S. Oh),전태현(T. H. Chun) 대한기계학회 2001 대한기계학회 춘추학술대회 Vol.2001 No.9
The computational fluid dynamics (CFD) method was used to determine an optimum design of hybrid mixing vane in a nuclear fuel bundle. The hybrid mixing vane is a new flow mixing device under development by Korea Atomic Energy Research Institute, which consists of two sets of primary and secondary vanes. Swirling flow and cross flow is primary mechanisms of coolant mixing in the fuel bundle. To maximize the coolant mixing by the hybrid vane, its size and vane angle must be optimized. Assuming an optimum size of the hybrid vane based on engineering judgment, the vane angles, defined as the angle bent from the axial flow direction, were varied from 30˚ to 40˚ and from 25˚ to 45˚ for the primary and secondary vanes, respectively. The swirl mixing increased as both the primary and secondary vane angle increases. The crossflow mixing appeared to increase as the primary vane angle increases and the secondary vane angle decreases. Turbulent mixing showed negligible dependence on the vane angles. Pressure drop due to the hybrid vane continually increased as the vane angle increases. The swirl and cross flow mixing factors were estimated from the predicted velocity distributions in the fuel bundle. The optimal vane angles are judged to be 40˚ and 25˚-35˚ for the primary and secondary hybrid vanes, respectively.
5X5 핵연료 봉다발에서 지지격자 가공 상태에 따른 압력손실 실험
박주용(J. Y. Park),오동석(D. S. Oh),장석규(S. K. Chang),인왕기(W. K. In) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10
In Korea Atomic Energy Research Institute(KAERI), it holds a patent right in the U.S.A, Korea, and Japan about development of fuel rod spacer grid for high performance of Pressurized Water Reactor(PWR). We performed experiments about hydraulic characteristics of the developed spacer grid. There were 4 types of spacer grid such as Plain Spacer Grid, Chamfering/Coining, spacer grid with mixing vane without Chamfering/Coining treatment(Mixing Vane), spacer grid with spot welding and chamfering/coining treatment(Mixing Vane_SC), and spacer grid with line welding in Mixing Vane_SC(Mixing Vane_LC). We compared pressure drop coefficient and found the pressure drop coefficient is greatly affected by chamfering/coining treatment and wasnt affected by welding procedure.
소듐고속로 핵연료집합체 측면 오리피스 주입구 난류유동의 전산유체역학 해석
인왕기(W. K. In),정영신(Y. S. Jeong,),이찬(C. Lee),신창환(C. H. Shin),오동석(D. S. Oh),전태현(T. H. Chun),천진식(J. S. Cheon) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11
A liquid sodium is used as coolant in sodium fast reactor(SFR). The reactor coolant pump supplies sodium coolant into a fuel assembly through the inlet chamber with a side orifice. Nine(9) orifices are used in the inlet chamber in three longitudinal and azimuthal directions. This CFD study investigates turbulent flow structure in the inlet chamber and estimate the pressure loss coefficient at the side orifice. The Reynolds number for this CFD simulation is 104, 105 and 2x105 based on the hydraulic diameter and bulk velocity in the orifice. Turbulence models used are the standard k-ε model, SAS-SST model, SSG Reynolds stress model and Large Eddy Simulation(LES). A unsteady flow simulation was performed to more accurately analyze the complex turbulent flow in the inlet chamber with the 9 side orifices. This paper presents the CFD predictions of turbulent flow distribution in the inlet chamber of SFR fuel assembly and the loss coefficient of side orifice.
가압경수로 이중냉각핵연료의 내측수로 막힘에 대한 전산유체역학 해석
인왕기(W.K. In),신창환(C.H. Shin),박주용(J.Y. Park),오동석(D.S. Oh),이치영(C.Y. Lee),전태현(T.H. Chun) 한국전산유체공학회 2011 한국전산유체공학회 학술대회논문집 Vol.2011 No.5
A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet(UO₂) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid fuel by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.
봉간격이 좁은 봉다발에서 압력손실의 실험 및 전산해석 평가
신창환(C.H. Shin),이치영(C.Y. Lee),박주용(J.Y. Park),오동석(D.S. Oh),인왕기(W.K. In) 한국전산유체공학회 2011 한국전산유체공학회 학술대회논문집 Vol.2011 No.11
A dual-cooled annular nuclear fuel has been introduced for a significant increase in reactor power. The KAERI has been researching the development of a dual-cooled annular fuel for a power increase in an optimized PWR in Korea, OPR-1000. The pitch-to-diameter ratio of the annular fuel assembly is decreased in order to maintain the fuel amount ratio within the same assembly size as the solid fuel assembly. In the tight lattice rod bundle, a pressure loss for the rod friction may be definitely different from that in the conventional solid fuel assembly. In this study, the friction loss of the 4x4 or 5x5 bare rod bundles without the obstacles such as the spacer grids is measured for the pitch-to-diameter 1.08 and 1.35, respectively. The measured results are compared with the general correlations for the conventional rod bundle. CFD studies are performed to investigate the friction pressure loss for the simulated single rod geometry and the spacer grid effects in the rod bundle is estimated in the view of a friction loss for a rod.