• 한국음향학회:학술대회논문집
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• 한국음향학회 2000년도 학술발표대회 논문집 제19권 2호
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• pp.231-234
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• 2000

In relation to a non-destructive evaluation of irradiation damage of micro-structure of interstitial, void and dislocation, the changes in the hysteresis loop and Barkhausen noise amplitude and the harmonics frequency due to neutron irradiation were measured and evaluated. The Mn-Mo-Ni low alloy steel of reactor pressure vessel was irradiated to a neutron fluence of $2.3\times10^{19}n/cm^2$ $(E\ge1MeV)$ at $288^{\circ}C.$The saturation magnetization of neutron irradiated metal did not change. Neutron irradiation caused the coercivity to increase, whereas susceptibility to decrease. The amplitude of Barkhausen noise parameters associated with the domain wall motion were decreased by neutron irradiation. The spectrum of Barkhausen noise was analyzed with an applied frequency of 4Hz and 8Hz, and a sampling time of 50 $\mu$ sec and 20 $\mu$ sec. The harmonic frequency of Joule effect shows 4Hz, 8Hz, 12Hz and 16Hz reflected from an unirradiated specimen. On the contrary, the harmonic frequency disappeared for the irradiated specimen. Harmonic frequency of induced voltage of sinusoidal magnetic field And Spectrum of Barkhausen noise on material is determined.

• 한국해양공학회지
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• 제8권2호
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• pp.173-184
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• 1994

원자력 압력용기강의 탄소성 파괴인성값 $J_IC$를 CT형 시험편을 이용하여 검토하였으며, 평활 시험편 및 측면홈 시험편의 두께는 각각 $B_O$=25.4mm, $B_N$=20.4mm 이다. 측면홈의 깊이는 19.7% 이며, 홈의 각도는 90 .deg.로 가공하였다. 탄소성 파괴인성시험은 ASTM E813-81과 JSME S001-81의 추천방법에 따라 실시하였다. 두 추천방법으로 실험한 결과 ASTM 방법에 의한 $J_IC$값이 과대평가됨으로써 부대조건에 만족되지 못하였지만 JSME방법은 만족되었다. 측면홈 시험편은 R고선법에 의한 ductile tearing의 결정이 평활 시험편보다 용이하였으며, 이에 따른 $J_IC$값의 정확성을 배가 할 수 있었다. 또한 임계 스트레치존 폭($SZW_C$)은 측면홈에 의한 높은 3축응력이 발생되어 평활시험편보다 적게 나타났으며, 이러한 복합적인 원인에 기인하여 스트레치존법에 의한 $J_IC$평가는 R곡선법에 의한 평가보다 약간 과대평가됨을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4022-4032
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• 2021

In this study, machine learning (ML) techniques were used to model surveillance test data of nuclear power plants from an international database of the ASTM E10.02 committee. Regression modeling was conducted using various techniques, including Cubist, XGBoost, and a support vector machine. The root mean square deviation of each ML model for the baseline dataset was less than that of the ASTM E900-15 nonlinear regression model. With respect to the interpolation, the ML methods provided excellent predictions with relatively few computations when applied to the given data range. The effect of the explanatory variables on the transition temperature shift (TTS) for the ML methods was analyzed, and the trends were slightly different from those for the ASTM E900-15 model. ML methods showed some weakness in the extrapolation of the fluence in comparison to the ASTM E900-15, while the Cubist method achieved an extrapolation to a certain extent. To achieve a more reliable prediction of the TTS, it was confirmed that advanced techniques should be considered for extrapolation when applying ML modeling.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 한국전산유체공학회지
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• 제10권3호
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.274-284
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• 1992

가압경수로 최적 열수력 분석용 전산코드인 CATHRE의 모델 평가를 위하여 가압경수로의 가상 냉각재 상실사고시 원자로 용기내의 유동현상을 모의한 1/15축소의 CREARE 실험을 모의 계산하였다. 이 실험에서 주요변수들은 비상노심 탱각재 주입량과 아냉정도 그리고 계통압력 및 노심에서 발생되는 증기유량이지만. 본 연구에서는 우선 Downcomer에서 역방향유동의 정성적 분석에 촛점을 맞추었다. 모의 계산 결과와 실험 결과를 비교할 때 정량적인 값 뿐 아니라 변화의 경향에서도 차이가 나타난 것은 주로 적절하지 못한 일부의 수치해석 모델과 상간의 계면마찰 때문으로 판단된다. 따라서 매개변수적 민감도 분석을 통하여 CATHARE 전산코드의‘VOLUME’에 접한 접합점에서 운동량 보존방정식의 상세연구 혹은 다차원 분석을 통해서 이 경우의 물리적 현상을 보다 현실적으로 나타낼 수 있다는 결론을 얻었다.

• 에너지공학
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• 제21권4호
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• pp.419-426
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• 2012

본 연구에서는 기기 스케일 2상 유동(Two-phase flow) 해석 코드 CUPID를 사용하여 CANDU 원자로의 칼란드리아 용기 내부 감속재의 열수력 거동을 분석하기 위한 사전연구를 수행하였다. 먼저, Stern 연구소에서 수행한 단상유동 실험 3종류를 이용하여 CUPID 코드를 검증하였다. 칼란드리아 관다발 영역 격자생성의 복잡성을 피하기 위하여 다공성 매질 모델을 해당 영역에 적용하였고, 다공성 매질 영역의 유동 저항은 실험에서 얻은 관계식을 이용하여 계산하도록 하였다. 계산결과, CUPID 코드는 칼란드리아 용기 내부의 강제 및 자연 대류의 혼합 유동 양식을 성공적으로 예측하였다. 다음으로 2상 유동이 발생하는 경우를 해석하였다. 이들 계산을 통해 CUPID 코드의 CANDU 원자로 감속재 해석 능력을 보였다. 또한, 국부 과냉각 여유도를 예측하는데 사용할 수 있는 유입유량 대비 칼란드리아 용기의 국부 최대 감속재 온도 그래프를 제시하였다.

• 대한기계학회논문집A
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• 제24권2호
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• pp.303-310
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• 2000

Fracture toughness of five different reactor pressure vessel steels was characterized in the transition temperature region by the ASTM E1921-97 standard method using Charpy-sized small specimens. T he predominant fracture mode of the tested steels was transgranular cleavage in the test conditions. A statistical analysis based on the Weibull distribution was applied to the interpretation of the scattered fracture toughness data. The size-dependence of the measured fracture toughness values was also well predicted by means of the Weibull probabilistic analysis. The measured fracture toughness transition curves followed the temperature-dependence of the ASTM master curve within the expected scatter bands. Therefore, the fracture toughness characteristics in the transition region could be described by a single parameter, so-called the reference temperature (T。), for a given steel. The determined reference temperatures of the tested materials could not be correlated with the conventional index temperatures from Charpy impact tests.

• 한국분말재료학회지
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• 제23권5호
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• pp.384-390
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• 2016

Tri-isotropic (TRISO) coatings on zirconia surrogate beads are deposited using a fluidized-bed vapor deposition (FB-CVD) method. The silicon carbide layer is particularly important among the coated layers because it acts as a miniature pressure vessel and a diffusion barrier to gaseous and metallic fission products in the TRISO-coated particles. In this study, we obtain a nearly stoichiometric composition in the SiC layer coated at $1400^{\circ}C$, $1500^{\circ}C$, and $1400^{\circ}C$ with 20 vol.% methyltrichlorosilane (MTS), However, the composition of the SiC layer coated at $1300-1350^{\circ}C$ shows a difference from the stoichiometric ratio (1:1). The density decreases remarkably with decreasing SiC deposition temperature because of the nanosized pores. The high density of the SiC layer (${\geq}3.19g/cm^2$) easily obtained at $1500^{\circ}C$ and $1400^{\circ}C$ with 20 vol.% MTS did not change at an annealing temperature of $1900^{\circ}C$, simulating the reactor operating temperature. The evaluation of the mechanical properties is limited because of the inaccurate values of hardness and Young's modulus measured by the nano-indentation method.