• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2005년도 연구개발 발표회 논문집
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• pp.244-249
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• 2005

Flow distribution and pressure drop analysis for an inlet plenum of a Pebble Bed Modular Reactor (PBMR) have been performed using Computational Fluid Dynamics. Three-dimensional Navier-Stokes equations have been solved in conjunction with $k-{\epsilon}$ model as a turbulence closure. Non-uniformity in flow distribution is assessed for the reference case and parametric studies have been performed for rising channels diameter, Reynolds number and angle between the inlet ports. Also, two different shapes of the inlet plenum namely, rectangular shape and oval shape, have been analysed. The relative flow mal-distribution parameter shows that the flow distribution in the rising channels for the reference case is strongly non-uniform. As the rising channels diameter decreases, the uniformity in the flow distribution as well as the pressure drop inside the inlet plenum increases. Reynolds number is found to have no effect on the flow distribution in the rising channels for both the shapes of the inlet plenum. The increase in angle between the inlet ports makes the flow distribution in the rising channels more uniform.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1669-1679
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• 2017

Background: Steam generator (SG) is one of the significant components in the nuclear steam supply system. A variety of SGs have been designed and used in nuclear reactor systems. Every SG has advantages and disadvantages. A brief account of some of the existing SG designs is presented in this study. A high surface to volume ratio of a SG is required in small modular reactors to occupy the least space. In this paper, performance improvement for SGs of integral small modular reactor is proposed. Aims/Methods: For this purpose, cross-grooved microfins have been incorporated on the inner surface of the helical tube to enhance heat transfer. The primary objective of this work is to investigate thermal-hydraulic behavior of the proposed improvements through modeling in RELAP5-3D. Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure). The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1244-1249
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• 2023

Recently, the Kingdom of Saudi Arabia (KSA) announced the development of first-of-a-kind(FOAK) and most advanced futuristic vertical city and named as 'The LINE'. The project will have zero carbon dioxide emissions and will be powered by clean energy sources. Therefore, a study was designed to understand which clean energy sources might be a better choice. Because of its nearly carbon-free footprint, nuclear energy may be a good choice. Nowadays, the development of very small modular reactors (vSMRs) is gaining attention due to many salient features such as cost efficiency and zero carbon emissions. These reactors are one step down to actual small modular reactors (SMRs) in terms of power and size. SMRs typically have a power range of 20 MWe to 300 MWe, while vSMRs have a power range of 1-20 MWe. Therefore, a study was conducted to discuss different vSMRs in terms of design, technology types, safety features, capabilities, potential, and economics. After conducting the comparative test and analysis, the fuel cycle modeling of optimal and suitable reactor was calculated. Furthermore, the levelized unit cost of electricity for each reactor was compared to determine the most suitable vSMR, which is then compared other generation SMRs to evaluate the cost variations per MWe in terms of size and operation. The main objective of the research was to identify the most cost effective and simple vSMR that can be easily installed and deployed.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.69-76
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• 2022

In order to prepare for a future nuclear market, research for developing floating small modular reactor has been initiated with the aim of differentiating it from large nuclear power plants such as distributed power, heat supply to remote communities and sea water desalination. Depending on the characteristics of the small modular reactor, it is necessary to design a plate and shell heat exchanger that can be manufactured smaller than the U-tube recirculation method. In this study, 12 cases are selected by changing the diameter of the heat plate, the thickness of the device body and the size of the stiffener. Finite element analysis is performed by setting the stress classification lines for the point at which deformation is expected under external pressure conditions for these analysis cases. For the basic design of the plate and shell heat exchanger, the optimal conditions are derived by analyzing the tendency of stress change in the device body and stiffener.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

• 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.

• 대한전기학회논문지:전력기술부문A
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• 제52권4호
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• pp.147-147
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• 2003

This paper describes the finite element analysis(FEA) for the design of electromagnet for Control Element Drive Mechanism(CEDM) in System-integrated Modular Advanced Reactor(SMART) and compared with the lifting power characteristics of prototype electromagnet. A thermal analysis was performed for the electromagnet. A model for the thermal analysis of the electromagnet was developed and theoretical bases for the model were established. It is important that the temperature of the electromagnet windings be maintained within the allowable limit of the insulation. since the electromagnet of CEDM is always supplied with current during the reactor operation. So the thermal analysis of the winding insulation which is composed of polyimide and air were performed by finite element method. As a result, it is shown that the characteristics of prototype electromagnet have a good agreement with the results of FEA. The thermal properties obtained here will be used as input for the optimization analysis of the electromagnet.

• 한국산학기술학회논문지
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• 제20권8호
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• pp.362-370
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• 2019

격납구조는 사고시 방사능 유출을 막기 위해 내압성능을 확보해야 하므로 소형 원자로용 격납구조에 모듈 방식을 적용하기 위해서는 내압성능의 분석이 필요하다. 따라서 소형 원자로용 모듈화 격납구조의 내압성능 분석을 위해 프리캐스트 콘크리트 모듈과 모듈 사이의 연결부 접촉면과 긴장재 배치를 고려한 FEM모델을 작성하고 정적해석을 수행한다. 이를 통해 모듈화 격납구조의 하중단계별 변위 및 응력의 변화특성을 분석한다. 그리고 변수 분석을 위해 선정된 각 변수가 모듈화 격납구조의 내압성능에 미치는 영향을 분석한다. 비교를 위해 일체화 격납구조의 내압성능도 함께 분석한다. FEM해석을 통한 변수 분석을 통해 긴장력 크기, 긴장재 배치 간격, 콘크리트 두께방향 긴장재 위치, 연결부 접촉면 마찰 계수 크기, 콘크리트 두께 등과 같은 변수 값의 범위가 제시되었다. 모듈화 격납구조의 모듈 간 접촉면에서 합성효과를 발생시켜주는 주요인자는 긴장재에 의한 긴장력과 연결부 접촉면의 마찰력이다. 일체화 격납구조 대비 추가적인 긴장재배치를 통해 긴장력을 증가시키면 모듈화 격납구조에서도 일체화 격납구조와 동등 수준의 내압성능을 확보할 수 있다.