• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1134-1145
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• 2021

A design parameter study is presented for the closed-loop type passive containment cooling system (PCCS) which is equipped with two heat exchangers: one installed at the inside of the containment and the other submerged in the water pool at the outside of the containment. A GOTHIC code model for PCCS performance analyses was set up and the design parameters such as the heat exchanger sizes, locations, and water pool tank volumes were analyzed to investigate the feasibility of installing this type of PCCS in PWRs like OPR-1000 being operated in Korea. We identified the size of the circulation loop and heat exchangers as major design parameters affecting the performance of PCCS. The analyses showed that the heat exchangers in the inside of the containment would be more influential on the heat removal capability of PCCS than that installed in the water pool at the outside of the containment. Hence, it was recommended to down-size the heat exchangers in the water pool to optimize PCCS without compromising its performance. Based on the parametric study, it was demonstrated that a closed-loop type PCCS could be designed sufficiently compact for installation in the available space within the containment of PWRs like OPR-1000.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1821-1829
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• 2023

The temperature distribution of the reactor pool under natural circulation induced by the RVACS operation was experimentally studied. According to the Bo' based similarity law, which could reproduce the temperature distribution of the working fluid under natural circulation, SINCRO-2D facility was designed based on the PGSFR. It was reduced to 1 : 25 in length scale, having water as a simulant of the sodium, which is the original working fluid. In general, temperature was stratified, however, effect of the natural circulation flow could be observed by the entrainment of the stratified temperature. Relative cooling contribution of the upper plenum (narrow gap) and lower plenum was approximately 0.2 and 0.8, respectively. In the range of decay heat from 0.2% to 1.0%, only the magnitude of the temperature was changed, while the normalized temperature maintained. Boundary temperature distribution change made a global temperature offset of the pool, without a significant local change. Therefore, the decay heat and cooling boundary condition had no significant effect on temperature distribution characteristics of the pool within the given range of the decay heat and boundary temperature distribution.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.302-310
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• 2011

A fluid in an enclosure can be heated by electric heating, chemical reaction, or fission heat. In order to remove the volumetric heat of the fluid, the walls surrounding the enclosure must be cooled. In this case, a natural convection occurs in the pool of the fluid, and it has a dominant role in heat transfer to the surrounding walls. It can augment the heat transfer rates tens to hundreds times larger than conductive heat transfer. The heat transfer by a natural convection in a regular shape such as a square cavity or semi-circular pool has been studied experimentally and numerically for many years. A pool of an inverted triangular shape with 10 degree inclined bottom walls has a good cooling performance because of enhanced boiling critical heat flux (CHF) compared to horizontal downward surface. The coolability of the pool is determined by comparing the thermal load from the pool and the maximum heat flux removable by cooling mechanism such as radiative or boiling heat transfer on the pool boundaries. In order to evaluate the pool coolability, it is important to correctly expect the thermal load by a natural convection heat transfer of the pool. In this study, turbulence models with modifications for buoyancy effect were validated for unsteady natural convections by volumetric heating. And natural convection in the triangular pool was evaluated by using the models.

• 한국전산유체공학회지
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• 제16권3호
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• pp.66-74
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• 2011

A fluid in an enclosure can be heated by electric heating, chemical reaction, or fission heat. In order to remove the volumetric heat of the fluid, the walls surrounding the enclosure must be cooled. In this case, a natural convection occurs in the pool of the fluid, and it has a dominant role in heat transfer to the surrounding walls. It can augment the heat transfer rates tens to hundreds times larger than conductive heat transfer. The heat transfer by a natural convection in a regular shape such as a square cavity or semi-circular pool has been studied experimentally and numerically for many years. A pool of an inverted triangular shape with 10 degree inclined bottom walls has a good cooling performance because of enhanced boiling critical heat flux (CHF) compared to horizontal downward surface. The coolability of the pool is determined by comparing the thermal load from the pool and the maximum heat flux removable by cooling mechanism such as radiative or boiling heat transfer on the pool boundaries. In order to evaluate the pool coolability, it is important to correctly expect the thermal load by a natural convection heat transfer of the pool. In this study, turbulence models with modifications for buoyancy effect were validated for unsteady natural convections by volumetric heating. And natural convection in the triangular pool was evaluated by using the models.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.45-60
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• 2006

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1990년도 한국자동제어학술회의논문집(국내학술편); KOEX, Seoul; 26-27 Oct. 1990
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• pp.367-371
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• 1990

In this paper, a real-time fault tolerant control system has been designed for the cooling system of the spent fuel pool storage. The fault tolerant control system consists of the fault detection part, the redundant actuator part(main and backup pumps) and the controller implemented on programmable. logic controller. This paper considers only the actuator fault whose detection is accomplished using Friedland's separated bias estimation method. This paper also shows the real-time experimental results from which it can be concluded that the designed fault tolerant control system exhibits satisfactory performance.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.309-314
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• 1996

Feasibility study has been accomplished to evaluate the effectiveness of the in-vessel core debris cooling through lower cavity flooding using two dimensional finite difference scheme. The volume of cerium pool and decay power rate generated in corium pool were evaluated as important parameters to the temperature distribution on the reactor vessel lower head through previous works. In this study, the corium volume based on the System 80＋ core structure and time dependent decay power rate are considered for feasibility evaluation. In addition, preliminary plans for the in-vessel core debris cooling through lower cavity flooding as severe accident management strategy, i.e. flooding timing, method and capacity, are suggested based on the result of the numerical study, international tendency related to in-vessel core debris cooling through lower cavity flooding.

• 대한기계학회논문집B
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• 제27권3호
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• pp.367-373
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• 2003

The present study investigates the fire suppression characteristics using a water mist fire suppression system. The fire extinguishing times are measured for various fire sources, fuel types, and different total flooding rates of water mist. Pool fire with hydrocabon fuel is successfully extinguished within a minute under the operating conditions of the water mist system. Two different regimes of the smoke layer cooling are observed, such as rapid and slow cooling processes. The regimes are divided by threshold time which is calculated with auto-correlation function. The threshold time for the initial cooling decreases with increasing water flow-rates and fire sources. These initial cooling effects play an important role in preventing the occurance of flashover fire by the initial fire suppression.

• Korean Chemical Engineering Research
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• 제47권4호
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• pp.446-452
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• 2009

최근 청정대체에너지로 각광을 받고 있는 dimethyl ether(DME)를 천연가스를 이용하여 직접 생산하는 1단계법의 고정층 촉매 반응기를 모사하였다. 1단계법의 고정층 반응기의 경우, 발열 반응인 메탄올 합성반응과 메탄올 탈수 반응이 동시에 일어나기 때문에 촉매의 비활성화를 일으킬 수 있는 hot spot의 발생을 막는 것이 가장 중요하다. 따라서 향후 상용공정에 적용 가능한 향류 냉각방식(count-current cooling system)과 포화액체 풀비등 방식(pool boiling system)을 적용하여 반응기 거동을 모사하고 이를 비교하였다. CO 전환율과 DME 생산성 면에서 향류 냉각방식이 더 효과적이었다. 하지만 포화비등액체 풀비등 방식이 더 작은 온도 범위에서 운전되어 hot spot(국소 고온점)의 가능성을 낮추고 더 안정적인 반응기 운전범위를 확보할 수 있었다.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.476-483
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• 2017

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of $0.22^{\circ}C/W$, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.