• Nuclear Engineering and Technology
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• 제39권4호
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• 공업화학
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• 제25권2호
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• pp.204-208
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• 2014

본 연구는 1차원 반응기 모델을 이용한 수치 시뮬레이션을 통해 수소투과량, 수소선택도, 사용된 촉매의 양, 급송흐름에서의 $H_2O/CO$ 조성비 및 Ar sweep gas가 막반응기(membrane reactor)에서의 수성가스전이반응의 성능에 미치는 영향을 분석하였다. 막반응기에서 평형상태보다 향상된 수소수율을 얻기 위해선 적어도 100 이상의 수소선택도를 가져야 함이 관찰되었으며, 수소투과량이 계속 증가될 경우에는 수소수율의 증가폭이 점차 감소됨이 보였다. 낮은 수소투과량의 경우에는 촉매량이 증가할수록 초기엔 증가된 CO 전환율을 보이다가 점차 그 증가폭이 감소되었으며, 높은 수소투과량의 경우에는 촉매의 양과 무관하게 높은 CO 전환율이 관찰되었다. 급송흐름에서의 $H_2O/CO$ 조성비가 1.5 이상인 경우엔 수소투과량이 막반응기에서의 CO 전환율에 미치는 영향이 미미하였고, 막반응기에서 평형상태보다 향상된 CO 전환율을 얻기 위해선 적어도 $6.7{\times}10^{-6}mol\;s^{-1}$ 의 Ar 몰유속이 필요함이 밝혀졌다.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.

• 전기의세계
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• 제26권6호
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• pp.66-71
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• 1977

In this paper a method is presented for the optimal control of a nuclear reactor at equilibrium state by use of a digital computer. Using the optimal control theory, we formulate the control problem of the reactor as a discrete-time linear regulator problem. A quadratic performance index is defined. The effects of choosing different performance index weighting matrices to the feedback gain matrix and reactor transient responses are studied for the deterministic optimal control with all state variables accessible to measurement.

• 한국환경보건학회지
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• 제22권4호
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• pp.10-15
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• 1996

Treatment of a synthetic wastewater was made for the purpose of evaluating the performance of a pilot scale of an IFBBR. The particles used in this experiment were circular shaped polystyrene whose density was lower than that of water. The reactor volume was 400 l and the SCOD removal efficiency was more than 90% at HRT of 5 hrs. The IFBBR was successfully operated for 5 months without any significant problems. And it whs certained that IFBBR had a potential capability for treating wastewater.

• Korean Journal of Chemical Engineering
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• 제35권11호
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• pp.2157-2163
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• 2018

Catalytic reactors have been essential for chemical engineering process, and different designs of reactors in multi-scales have been previously studied. Computational fluid dynamics (CFD) utilized in reactor designs have been gaining interest due to its cost-effective advantage in designing the actual reactors before its construction. In this work, butadiene synthesis via oxidative dehydrogenation (ODH) of n-butene using tubular reactor was used as a case study in the CFD model. The effects of coolant and reactor diameter were investigated in assessing the reactor performance. Based on the results of the CFD model, the conversion and selectivity were 86.5% and 59.5% respectively in a fixed bed reactor under adiabatic condition. When coolants were used in a tubular reactor, reactor temperature profiles showed that solar salt had lower temperature gradients inside the reactor than the cooling water. Furthermore, higher conversion (90.9%) and selectivity (90.5%) were observed for solar salt as compared to the cooling water (88.4% for conversion and 86.3% for selectivity). Meanwhile, reducing the reactor diameter resulted in smaller temperature gradients with higher conversion and selectivity.

• 한국가시화정보학회지
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• 제9권4호
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• pp.54-62
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• 2011

KAERI has developed a KALIMER-600 which is a pool-type sodium-cooled fast reactor with a 600MWe electric generation capacity. For a SFR development, one of the main topics is an enhancement of the reactor system safety. Therefore, we have a long-term plan to design the large sodium experimental facility to evaluate the reactor safety and component performance. In order to extrapolate a thermal hydraulic phenomena in a large sodium reactor, the thermal hydraulics phenomena is under investigation in a 1/$10^{th}$ water-simulant facility for the KALIMER-600. In this paper, we shortly described the experimental facility setup and the measurement of the isothermal global flow behavior. For the flow field measurement, the PIV method was used in a transparent Plexiglas reactor vessel model at around $20^{\circ}C$ water condition.

• 대한기계학회논문집B
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• 제30권8호
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• pp.818-824
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• 2006

The hydrodynamic performance analysis of an axial-flow main coolant pump for the modular nuclear reactor has been carried out using a commercial computational fluid dynamics (CFD) software. The prediction capability of the CFD software adopted in the present study was validated in comparison with the experimental data. Predicted performance curves agree satisfactorily well with the experimental results for the main coolant pump over the normal operating range. π Ie prediction method presented herein can be used effectively as a tool for the hydrodynamic design optimization and assist the understanding of the operational characteristics of general purpose axial-flow pumps.

• 한국수소및신에너지학회논문집
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• 제34권3호
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• pp.316-326
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• 2023

Development of carbon-neutral fuel production technologies to solve climate change issues is progressing worldwide. Among them, methane can be produced through the synthesis of hydrogen produced by renewable energy and carbon dioxide captured through a CO₂ methanation reaction, and the fuel produced in this way is called synthetic methane or e-methane. The CO₂ methanation reaction can be conducted via biological or thermochemical methods. In this study, a 30 Nm³/h thermochemical CO₂ methanation process consisting of an isothermal reactor and an adiabatic reactor was used. The CO₂ conversion rate and methane concentration according to the temperature measurement results at the center and outside of the adiabatic reactor were analyzed. The gas flow into the adiabatic reactor was found to reach equilibrium after about 1.10 seconds or more by evaluating the residence time. Furthermore, experimental and analysis results were compared to evaluate performance of the reactor.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.