• 기계저널
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• 제50권3호
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• pp.28-31
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• 2010

이 글에서는 제4세대(Generation-IV) 원자로시스템의 자원활용 측면에서 핵연료 주기와 관련하여 새롭게 부각되고 있는 소듐냉각고속로(SFR: Sodium-cooled Fast Reactor)의 개발 목적 및 설계 특성을 기술하고 원자로 구조관점에서 가압경수로(PWR)와 비교 설명한다.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2011년도 춘계학술대회 논문집
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• pp.889-890
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• 2011

• Journal of Radiation Protection and Research
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• 제32권4호
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• pp.168-177
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• 2007

본 논문에서는 노심설계 단계에서 선정된 다양한 노심 장전모형 중에서 압력용기 중성자 조사취화 관점에서 가장 최적의 노심 장전모형을 선정할 수 있도록 신속하게 압력용기 취약위치에 대한 속중성자속을 예측할 수 있는 방법을 제시하였다. 인공신경회로망 기법을 통해 노심 반경방향 및 축방향 출력분포만을 이용하여 압력용기내벽 취약위치에서의 중성자 스펙트럼을 신속하게 평가할 수 있도록 중성자속 가중치를 생산하였고 데이터베이스를 구축하였다. 이 방법은 중성자 수송코드를 이용한 수송계산을 직접 수행하지 않고도 신속하게 압력용기 위치에서의 중성자 조사환경을 평가할 수 있으며 소송코드 결과와 비교하여 상대오차 3.4%이내의 정확도를 보였다.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.107-112
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• 2014

The Code of Federal Regulations, Title 10, Part 50, Appendix H requires surveillance program for reactor pressure vessel(RPV) that the peak neutron fluence at the end of the design life of the vessel will exceed $1.0E+17n/cm^2$ (E>1.0MeV). 2D/1D Synthesis method based on DORT 3.1 transport calculation code has been widely used to determine fast neutron(E>1.0MeV) fluence exposure to RPV in the beltline region. RAPTOR-M3G(RApid Parallel Transport Of Radiation-Multiple 3D Geometries) performing full 3D transport calculation was developed by Westinghouse and KRIST(Korea Reactor Integrity Surveillance Technology) and applied for the evaluations of In-Vessel and Ex-Vessel neutron dosimetry. The reaction rates from measurement and calculation were compared and the results show good agreements each other.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.243-256
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• 2011

This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution.

• 한국유체기계학회 논문집
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• 제18권6호
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• pp.5-11
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• 2015

In a pressurized water reactor (PWR), control rod assembly (CRA) falls into the guide tubes of a fuel assembly due to gravity for scram. Various theoretical approaches and numerical analyses have been performed because its shape is simple and its design was completely developed several decades ago. A control rod assembly for a sodium-cooled faster reactor (SFR) which is geometrically more complicated is being actively developed in Korea nowadays. Drop time and impact velocity of a CRA are important parameters with respect to reactivity insertion time and the mechanical robustness of a CRA and a guide duct. In this paper, computational method considering simultaneously the equation of motion for rigid body and the Navier-Stokes equations for fluid is suggested and verified by comparison with theoretical analysis results. Through this valuable CFD analysis method, drop time and impact velocity of initially designed SFR CRA are evaluated before performing scram tests with it.

• 한국전산유체공학회지
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• 제16권3호
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• pp.95-103
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• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1813-1824
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• 2022

This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from -0.3$ to -0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.