• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2004년도 추계학술대회 논문집
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• pp.1327-1330
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• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.

• 방사성폐기물학회지
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• 제16권3호
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• pp.359-367
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• 2018

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

• 상하수도학회지
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• 제21권1호
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• pp.83-89
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• 2007

A biological anoxic-aerobic reactor filled with porous media was operated in lab scale for the advanced wastewater treatment. The experiments were conducted for 6 months with three HRTs (4, 6, 8hr) and temperature of $23{\sim}25^{\circ}C$. Some other experimental conditions were as follows; nitrification reactor (MLSS 4,500mg/L, DO 3.3mg/L, $23{\sim}28^{\circ}C$), denitrification reactor(MLSS 8,000mg/L, ORP -100mV, Temp.$19{\sim}23^{\circ}C$). Average removal efficiencies of SS, $BOD_5$, $COD_{Cr}$, T-N, and T-P were 97.8%, 95.5%, 94.5%, 80.2%, and 60.6%, respectively. The reactor filled with porosity media showed stable removal capacity for organics and nutrients. Fast and complete nitrification and denitrification were accomplished. Maintaining high MLSS with porous media in the nitrification and denitrification reactor appears to enhance the nitrogen removal process. For the higher T-P removal, some coagulant addition process will be needed.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

• 대한기계학회논문집A
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• 제34권9호
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• pp.1217-1225
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• 2010

소듐을 냉각재로 사용하는 고속로의 증기발생기에서는 소듐과 물의 화학적 반응을 최소화하는 것이 중요한 문제이다. 소듐과 물의 반응 가능성을 줄여 증기발생기의 신뢰성을 향상시키기 위한 한가지 방안으로 이중벽관을 전열관으로 사용하는 증기발생기를 개발하고 있다. 이 증기발생기에서 중요한 현안은 이중벽관에서의 열전달 성능을 향상시키는 문제와 원자로 운전 중에 소듐과 물 반응사고가 일어나기 전에 전열관의 파손을 감지하는 기술을 개발하는 것이다. 이 논문에서는 이 현안을 극복할 수 있는 방안을 제시하였고, 이 기술을 활용하여 증기발생기의 개념을 설계하였다. 또한 이 개념에 적용되는 이중벽관을 설계 및 예비 제작하여 기계적 시험을 수행하였다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• 제21권2호
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• pp.99-110
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• 1989

원자로 지논 농도의 최적 제어는 Linear Quadratic Regulator Problem이다. 지논 농도와 아이오다인 농도는 측정할 수 없기 때문에 최적 제어를 수행하기 위해서는 측정할 수 없는 상태 변수를 예측하는 것이 필요하다. 본 연구에서 사용된 예측방법은 Luenberger Observer를 기초로 했다. 원자로 상태 방정식은 빠른 상태 방정식(중성자 속, 핵연료 및 냉각재 온도)과 느린 상태 방정식(아이오다인, 지논)의 상호작용에 의해 Stiffness 문제가 발생되는데 이러한 시스템을 "Singularly Perturbed System"이라 한다. Stiffness문제를 해결하기 위해서 원 시스템을 느린 시스템과 빠른 시스템의 두 개의 모드로 나누는 Singular Perturbation Method를 사용한다. 예측기Observer를 이용한 원 시스템의 제어기는 느린 시스템과 빠른 시스템에 대한 분리된 예측기와 제어기의 설계에 의해 결정되어진다. 특히 원자로 상태 방정식에서는 빠른 모드는 빨리 사라지게 되므로 단지 느린 시스템에 대해서만 예측기를 설계하면 된다. 컴퓨터시뮬레이션을 통한 시험 결과는 원자로의 지논 진동은 Singular Perturbation Method와 예측기를 이용해서 거의 정확하게 효과적으로 짧은 시간내에 제어할 수 있음을 알았다.수 있음을 알았다.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.323-334
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• 2013

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.897-907
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• 2020

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.