• Journal of Radiation Protection and Research
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• 제36권1호
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• pp.35-43
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• 2011

The PHENIX Experiment is the largest of the four experiments that have taken data at the Relativistic Heavy Ion Collider. PHENIX, the Pioneering High Energy Nuclear Interaction eXperiment, is designed specifically to measure direct probes of the collisions such as electrons, muons, and photons. The primary goal of PHENIX is to discover and study a new state of matter called the Quark-Gluon Plasma. Among many particles, muons coming from W-boson decay gives us key information to analyze the spin of proton. Resistive plate chambers are proposed as a suitable solution as a muon trigger because of their fast response and good time resolution, flexibility in signal readout, robustness and the relatively low cost of production. The RPC detectors for upgrade were assembled and their performances were evaluated. The procedure to make the detectors better was optimized and described in detail in this thesis. The code based on ROOT was written and by using this the performance of the detectors made was evaluated, and all of the modules for north muon arm met the criteria and installation at PHENIX completed in November 2009. As RPC detectors that we made showed fast response, capacity of covering wide area with a resonable price and good spatial resolution, this will give the opportunity for applications, such as diagnosis and customs inspection system.

• 원자력산업
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• 5_6호통권7호
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• pp.34-42
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• 1982

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.16-24
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• 1985

풀형 고속증식로에서의 과도 현상을 모사할 수 있는 전산 모델이 개발되었다. 이 전산 모델 SIM-FARP는 어떠한 펌프로의 전원 상실사고나 완전한 강제냉각 상실사고, 그리고 자연순환 과정 등을 모사할 수 있는 Fast Running Computer Code이다. 이에 따라 8개의 지배방정식이 유도되었으며, 이8개의 미분 방정식을 풀기 위해 Runge-Kutta의 수치해석방법이 사용되었다. 개발된 전산 프로그램은 두 가지 예제에 적용되었는데 이는 Super-Phenix-I에서의 펌프에의 전원상실사고 및 원자로가 정지되지 않는 상태에서의 외부전원 상실사고이다.

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.78-84
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• 1986

일차섭동이론을 이용하여 물질밀도 감도 계수의 표현식을 유도하였다. Super-Phenix I 평형노심의 초기상태를 기준계로 택했으며 유효중배계수를 계의 응답함수로 정의했다. 볼츠만 연산자의 구성요 소인 물질밀도로 표현되는 핵연료의 농축도와 실효밀도를 입력변화로 선정했다. 위 계산을 수행하는데 전산코드시스템 (KAERI-26군 단면적 library/1DX/2DB/PERT-V)가 사용되었다. 핵연료 농축도의 유효증배계수에 대한 감도계수는 4.576로 계산되었으며, 핵연료 실효밀도의 감도 계수는 0.0756으로 계산되었다. 본 연구는 감도해석법이 대형전산코드를 이용한 직접반복계산법에 비해 계산시간의 단축과 아울러 많은 정보를 준다는 것을 보여준다.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• 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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• 제16권2호
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• pp.70-79
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• 1984

대형 고속증식로용 핵계산 체제(KAERI-26군 단면적 library/1DX/2DB)를 이용하여 SUPER-PHE-NIX I 평형 노심의 초기상태에서의 diagrid 및 Pad팽창에 대한 반응도 계수를 계산, 검토하였다. 노심은 R-Z등가 model로 묘사하고, 2차원 다군 확산 이론 전산코드인 2DB를 사용하여 임계도를 계산하였다. 기초계산으로서 반경방향 및 축방향 균일 괭창에 대한 반응도 계산과 노심구성물질의 원자수 밀도 변화와 노심체적 변화에 대한 반응도 변화량계산을 수행하였다. 균일 팽창으로 고려한 diagrid팽창에 대한 반응도 계수는 -0.553pcm/mil로 계산되었다. 한편 반응도의 온도계수는 -1.0766pcm/$^{\circ}C$로 환산되어 프랑스 발표치 -1.09pcm/$^{\circ}C$와 1.2%오차내로 일치하였다. Diagrid 팽창효과 졔산방법을 활용하여 노심의 불균일 팽창을 유발하는 pad팽창에 대한 반응도 계수 계산을 수행한 바 매우 유용함을 알았다. Pad팽창에 대한 반응도 계수는 평균팽창 model의 경우 -0.2743pcm/mi1, pancake집적 model의 경우 -0.2786pcm/mi1로 각각 계산되었다. 또한 자유팽창 노심에서의 온도변화에 따른 pad 팽창에 대한 반응도 계수는 각 model에 대하여 -0.5766pcm/$^{\circ}C$, -0.5858pcm/$^{\circ}C$로 계산되어 프랑스 발표치 -0.574pcm/$^{\circ}C$와 2% 오차내로 일치하였다.