• 한국폐기물자원순환학회지
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• 제34권7호
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• pp.685-696
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• 2017

This study focuses on computational particle fluid dynamics (CPFD) modeling for the fast pyrolysis of biomass in a conical spouted bed reactor. The CPFD simulation was conducted to understand the hydrodynamics, heat transfer, and biomass fast pyrolysis reaction of the conical spouted bed reactor and the multiphase-particle in cell (MP-PIC) model was used to investigate the fast pyrolysis of biomass in a conical spouted bed reactor. A two-stage semi-global kinetics model was applied to model the fast pyrolysis reaction of biomass and the commercial code (Barracuda) was used in simulations. The temperature of solid particles in a conical spouted bed reactor showed a uniform temperature distribution along the reactor height. The yield of fast pyrolysis products from the simulation was compared with the experimental data; the yield of fast pyrolysis products was 74.1wt.% tar, 17.4wt.% gas, and 8.5wt.% char. The comparison of experimental measurements and model predictions shows the model's accuracy. The CPFD simulation results had great potential to aid the future design and optimization of the fast pyrolysis process for biomass.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1994년도 추계학술발표회 초록집
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• pp.146-149
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• 1994

A nonlinear dynamic optimization has been performed for reactor power control system of CANDU 6 nuclear power plant considering xenon, fuel and moderator temperature feedback effects. Integral-of-Time-multiplied Absolute-Error (ITAE) criterion has been used as a performance index of the system behavior. Optimum controller gain are found by searching algorithm of Sequential Quadratic Programming (SQP). System models are referenced from most recent literatures. Signal flow network construction and optimization have been done by using commercial computer software package.

• 한국가스학회지
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• 제20권6호
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• pp.95-101
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• 2016

본 연구에서는 다양한 수소 생산 방법 중 하나인 천연가스 수증기 개질반응(natural gas steam reforming reaction)에 대해 일반적인 충전층반응기와 반응기와 수소분리기가 결합된 새로운 형태의 분리막 반응기에서의 성능에 대한 비교분석을 수행하였다. Xu 와 Froment에 의해 기존에 발표된 실험결과를 바탕으로 상업용 화학공정모사기인 Aspen $HYSYS^{(R)}$ 모델이 개발되었으며, 반응온도, $H_2$ 투과량, Ar 유량 등이 분리막 반응기에서의 반응물의 전환율 및 $H_2$ 수율 향상도에 미치는 영향에 대해 분석한 결과 분리막 반응기에서 보다 많은 양의 수소수율 및 메탄전환율이 확인되었다. 더 나아가, 전체 시스템에서 필요로 하는 열량을 공급하기 위해 요구되는 천연가스의 양에 초점을 맞춰 분리막 반응기에서의 원가절감 가능성을 평가한 결과, 분리막 반응기에서 10.94%의 원가절감이 관찰되었다.

• 대한기계학회논문집B
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• 제37권10호
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• pp.961-969
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• 2013

본 연구에서는 상용 전산유체역학 소프트웨어인 ANSYS CFX V.13을 사용하여 정상상태, 비압축성, 등온으로 가정된 엇갈림 관 다발 및 일렬 관 다발 주위의 난류유동을 계산하였다. 계산 결과는 전산유체역학 우수사례 지침에 근거하여 격자크기, 대류항 차분법의 정확도 및 난류모델에 대한 민감도 연구에 활용되었고 실험 결과와 정량적으로 비교함으로써 우수사례 지침의 적용성을 평가하였다. 결론적으로 전산유체역학 우수사례 지침이 관 다발 유동 분야에서 상용 전산유체역학 소프트웨어의 예측성능 향상을 반드시 보증하지 않음을 확인하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.1001-1007
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• 1995

A Comprehensive Vibration Assessment Program (CVAP) has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibrations prior to commercial operation. The theoretical evidence for the structural integrity of the reactor internals and the basis for measurement and inspection are provided by the analysis. Flow induced hydraulic loads and reactor internals vibration response data were measured during pre-core hot functional testing in YGN 4 site. Also, the critical areas in the reactor internals were inspected visually to check any existence of structural abnormality before and after the pre-core hot functional testing. Then, the measured data have been analyzed and compared with the predicted data by analysis. The measured stresses are less than the predicted values and the allowable limits. It is concluded that the vibration response of the reactor internals due to the flow induced vibration under normal operation is acceptable for long term operation.

• Nuclear Engineering and Technology
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• 제47권5호
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• pp.534-545
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• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.