• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 제36회 하계학술대회 논문집 B
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• pp.1008-1010
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• 2005

In the examination of steam generator(SG) tube in nuclear power plant, eddy current testing probes play an important role in detecting the defects. Bobbin probe and MRPC probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary MRPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it has excellent detection capability for small cracks, which is hardly detected by bobbin probe. In this paper, for the accurate analysis of experimental ECT signals, construction of MRPC probe signals database according to the variation of defect size is the main purpose. Using 3-D finite element method, ECT signals are analyzed, and signals analysis add according to frequency ingredient. The results, which are analysis and characteristics ion of electromagnetism simulation signals, is databased.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.640-647
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• 2023

Material corrosion in nuclear power plant (NPP) is not controlled only by amine injection but also by ion exchange (IX) which is the best option to remove trace Na⁺. This study was conducted to understand the Na⁺ leakage characteristics of IX beds packed with ethanolamine-form (ETAH-form) and hydrogen-form (H-form) resins in the simulated water-steam cycle in terms of intrinsic behaviors of four kinds of cation-exchange resins through ASTM test and Vanselow mass action modeling. Na⁺ was inappreciably escaped throughout the channel created in resin layer. Na⁺ leakage from IX bed was non-linearly raised because of its decreasing selectivity with increasing Na⁺ capture and with increasing the fraction of ETAH-form resin. Na⁺ did not reach the breakthrough earlier than ETAH⁺ and NH₄⁺ due to the increased selectivity of Na⁺ to the cation-exchange resin (H⁺ < ETAH⁺ < NH₄⁺ ≪ Na⁺) at the feed composition. Na⁺ leakage from the resin bed filled with small particles was decreased due to the enhanced dynamic IX processes, regardless of its low selectivity. Thus, the particle size is a predominant factor among intrinsic properties of IX resin to reduce Na⁺ leakage from the condensate polishing plant (CPP) in NPPs.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Journal of Welding and Joining
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• 제34권3호
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• pp.52-60
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• 2016

Recently nickel based superalloys are extensively being regarded as the materials for the steam turbine parts for hyper super critical (HSC) power plants working at the temperature over $700^{\circ}C$, since the materials have excellent strength and corrosion resistance in high temperature. In this paper, alloy 617 of solution strengthened material and alloy 263 of ${\gamma}^{\prime}$-precipitation strengthened material were prepared as the testing materials for HSC plants each other. Post weld heat treatment (PWHT) was conducted with the gas tungsten arc (GTA) welded specimens. The microstructure of the base metals and weld metals were investigated with Electron Probe Micro-Analysis (EPMA) and Scanning Transmission Electron Microscope (STEM). The experimental results revealed that Ti-Mo carbides were formed in both of the base metals and segregation of Co and Mo in both of the weld metals before PWHT and PWHT leaded to precipitation of various carbides such as Mo carbides in the specimens. Furthermore, fine ${\gamma}^{\prime}$ particles, that were not precipitated in the specimens before PWHT, were observed in base metal as well as in the weld metal of alloy 263 after PWHT.

• 비파괴검사학회지
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• 제20권1호
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• pp.10-17
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• 2000

화력발전소 보일러의 주증기관, 헤더, 스팀드럼 등과 같은 주요 고온배관설비에서 발생하는 크리프 손상을 측정하는 비파괴적 측정방법에는 레프리카, 전기저항법 및 경도법 둥이 적용되고 있으나, 이들 방법들은 측정절차 및 준비가 복잡할 뿐만 아니라 접근이 가능한 설비표면에만 적용되는 제한점을 가지고 있다. 따라서 본 논문은 이들 종래의 방법을 신뢰성 있고 정량적인 초음파 비파괴평가법으로 보완 및 적용을 위하여, 실제 고온배관의 운전조건을 모의하여 수행한 크리프 인공열화실험 및 이들 크리프손상재에 대한 초음파실험을 통한 주파수분석 연구로서, 크리프손상 상태별 초음파 신호 분류를 위해 초음파신호의 각종 주파수특성을 평가하였다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• 전자공학회논문지S
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• 제35S권2호
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• pp.137-145
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• 1998

In this paper, we propose a tube-hole center detection vision algorithm verifying the position of a tele-controlled robot and providing visual information for increasing reliability and efficiency in the diagnosis of steam generator (SG) tubes in nuclear power plant. A tele-controlled robot plays a role in carrying the probe used in inspecting the integrity of SG tubes. Thus accurately locating a tele-controlled robot on the desired tube-hole center is important issue for reliability of inspection. To do this work, we have to find the tube-hole center locations from the input image. At first, we apply the three-class segmentation method modified for this application. WE extract minimum bounding rectangles (MBRs) in the theresholded binary image. Second, for discriminating between MBR by tube and MBR by noise, we introduce the MBR rejection rules as knowledge-based rule set. MBRs are divided into the very dark region MBRs and the very bright region MBRs. In order to describe the region of complete tube-hole, the MBRs need a process of pairing each other. We then can find the tube-hole center from the paired MBR. For more accurately finding the tube-hole center in several sequential images, the centers of some frames need to be averaged. We tested the performance of our method using hundreds of real images.

• 에너지공학
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• 제12권2호
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• pp.145-153
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• 2003

석탄이용 무공해 발전공정의 핵심기술인 탄화공정을 개발하기 위하여, 본 공정의 주반응인 메탄 수증기 개질에 대한 특성을 조사하였다. 개질촉매를 이용한 메탄수증기 개질에서는 공간속도, 수증기/탄소-비, 압력에 대한 영향을 조사하였다. 공간속도 7,000$hr^{-1}$ 이하에서 평형 전화율을 얻었다. 혼성반응으로 구성된 탄화공정 중 메탄 수증기 개질 반응조건인 700~80$0^{\circ}C$, 수증기/탄소-비 2.5~3에서 생성물 조성분포는 상압에서 수소 75~78%, 이산화탄소 8~10%, 1~30기압에서 수소 60~78%, 이산화탄소 9~11%를 얻었다.