• Proceedings of the KIEE Conference
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• 2006.07b
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• pp.1159-1160
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• 2006

As the grid-connected photovoltaic power conditioning systems (PVPCS) are installed in many residential areas, this has raised potential problems of network protection on electrical power system. One of the numerous problems is an Island phenomenon. There has been an argument that because the probability of islanding is extremely low it may be a non-issue in practice. However, there are three counter-arguments: First, the low probability of islanding is based on the assumption of 100% power matching between the PVPCS and the islanded local loads. In fact, an island can be easily formed even without 100% power matching (the power mismatch could be up to 30% if only traditional protections are used, e.g. under/over voltage/frequency). The 30% power-mismatch condition will drastically increase the islanding probability. Second, even with a larger power mismatch, the time for voltage or frequency to deviate sufficient to cause a trip, plus the time required to execute the trip (particularly if conventional switchgear is required to operate), can easily be greater than the typical re-close time on the distribution circuit. And, third, the low-probability argument is based on the study of PVPCS. Especially, if the output power of PVPCS equals to power consumption of local loads, it is very difficult for the PVPCS to sustain the voltage and frequency in an island. Unintentional islanding of PVPCS may result in power-quality issues, interference to grid-protection devices, equipment damage, and even personnel safety hazards. So the verification of anti-islanding performance is strongly needed. In this paper, the authors propose the improved RPV method through considering power quality and anti-islanding capacity of grid-connected three-phase PVPCS in IEEE Std 1547 ("Standard for Interconnecting Distributed Resources to Electric Power Systems"). And the simulation and experimental results are verified.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.533-544
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• 2016

This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap) have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

• Corrosion Science and Technology
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• v.20 no.5
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.47 no.2
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• pp.114-121
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• 2019

Tailless aircraft have advantages of low visibility compared to conventional aircraft, but drawback of poor stability as well which makes designing controller difficult. The controller design is more difficult, especially when the center of gravity moves due to store release or fuel consumption during flight. In this paper, an L1 adaptive controller is proposed as a way to overcome these problems. The reliability and performance of the controllers were verified by non-linear simulations. RPV Flying Quality Design criteria were used for design criteria. Using the simulation, it is shown that the adaptive controller maintains stability of the unmanned aircraft for sudden large change in the inertial properties. It is also shown that the calculation burden can be reduced when it is used with the gain scheduling method.

• Korean Journal of Metals and Materials
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• v.46 no.12
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• pp.771-779
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• 2008

An effective way of increasing the strength and fracture toughness of reactor pressure vessel steels is to change the material specification from that of Mn-Mo-Ni low alloy steel(SA508 Gr.3) to Ni-Mo-Cr low alloy steel(SA508 Gr.4N). In this study, we evaluate the effects of alloying elements on the microstructural characteristics of Ni-Mo-Cr low alloy steel. The changes in the stable phase of the SA508 Gr.4N low alloy steel with alloying elements were evaluated by means of a thermodynamic calculation conducted with the software ThermoCalc. The changes were then compared with the observed microstructural results. The calculation of Ni-Mo-Cr low alloy steels confirms that the ferrite formation temperature decreases as the Ni content increases because of the austenite stabilization effect. Consequently, in the microscopic observation, the lath martensitic structure becomes finer as the Ni content increases. However, Ni does not affect the carbide phases such as $M_{23}C_6 $ and $M_7C_3$. When the Cr content decreases, the carbide phases become unstable and carbide coarsening can be observed. With an increase in the Mo content, the $M_2C$ phase becomes stable instead of the $M_7C_3$ phase. This behavior is also observed in TEM. From the calculation results and the observation results of the microstructure, the thermodynamic calculation can be used to predict the precipitation behavior.

• Journal of the Korean Society for Nondestructive Testing
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• v.9 no.1
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• pp.106-110
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• 1989

Bolting degradation problems in primary coolant pressure boundary applications have become a major concern in the nuclear industry. In the bolts concerned, the failure mechanism was either corrosion wastage(loss of bolt diameter) or stress-corrosion cracking.(3) Here the manual ultrasonic testing of RPV(Reactor Pressure Vessel) and RCP(Reactor Coolant Pump) stud has been performed. But it is difficult to detect indications because examiner can not exactly control the rotation angle and can not distinguish the indication from signals of bolt. In many cases, the critical sizes of damage depth are very small(1-2 mm order). At critical size, the crack tends to propagatecompletly through the bolt under stress, Resulting in total fracture.(3) Automatic stud scanner for studs(bolts) was developed because the precise measurement of bolt diameter is required in this circumstance. By use of this scanner, the rotation angle of probe was exactly controlled and the exposure time of radiations was reduced.

• Journal of the Korean Society for Nondestructive Testing
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• v.16 no.1
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• pp.46-51
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• 1996

원자로는 압력용기 및 내부 구조물로 구분되어 있다. 내부 구조물들의 경련 열화 현상에 따라 결함이 많이 발생하여 이에 대한 초음파검사가 요구되고 있다. 따라서 원자로 내부 구조물에 대한 초음파검사 현황 및 각각 구조물들의 검사 원리를 기술하였다. 특히 원자로 내부 구조물 중 CRDM, core baffle bolt, core barrel bolt, CRGT-support pin 및 fuel alignment pin에 대한 유럽 및 독일을 중심으로 한 검사 현황 및 검사방법을 간략하게 기술하였다. 이 기술에 대한 지침안(guideline)이 독일, 프랑스, 일본을 중심으로 하여 마련되고 있다.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.25-40
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• 2016

The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. In this study, the SE methodology is used to design a nondestructive inspection system for Bottom Mounted Instrumentation (BMI) nozzles. We developed a system that enables nondestructive inspection of BMI nozzles during regular refueling outage without removing the reactor internals. A special ultrasonic (UT) probe is introduced to scan and detect cracks within the weld region of the nozzle. A 3D model of the inspection structure system was developed along with the reactor pressure vessel (RPV) and internals which permits a virtual 3D simulation of the operation to check the design concept and effectiveness of the system and to provide a good visualization of the system. This approach allows for a virtual walk through to verify the proposed BMI nozzle inspection system.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.1578-1583
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• 2003

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the coolability in case of the external reactor vessel cooling (ERVC). All the 4 tests have been performed using Alumina iron thermite melt as a corium simulant. Due to the limited steam venting through the insulation, steam binding occurred inside the annulus in the KSNP case simulation. On the contrary, in the tests which were performed for simulating the APR1400 insulation design, sufficient water ingression and steam venting through the insulation lead to effective cool down of the vessel characterized by nucleate boiling. It could be found from the experimental results that modification of the insulation design allowing sufficient ventilation could increase the positive effects of the external reactor vessel cooling.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.771-776
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• 1998

TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10$^{18}$ n/$\textrm{cm}^2$. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60cm below the midplane at zero degree.