• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.151-156
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• 2001

Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1992년도 가을 학술발표회 논문집
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• pp.67-72
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• 1992

The heat of hydration of cement causes the internal temperature rise at early age, particulary in massive concrete structures such as a footing of nuclear reactor building or a dam. As the result of the temperature rise and restraint of foundation, the thermal stress may induce cracks in concrete. Therefore, the prediction of the thermal stress is very important in the design and construction stages in order to control the cracks developed in massive concrete structures. And, in case of young concrete, creep effect by the temperature load is larger than That of old concrete. Thus the effect of creep must be considered for checking the cracks, serviceability, durability and leakage. This study is composed of two items. The first, it is to develop a finite element program which is capable of simulating the temperature history in mass concrete. The second, when the thermal stress of mass concrete structures considering creep is calculated by using the modified elastic modulus due to the inner temperature change. It is shown that the analytical results of this study is in comparably good agreement with JCI's analytical results.

• 대한인간공학회지
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• 제33권1호
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• pp.27-37
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• 2014

Objective: This study aims to improve human performance by analyzing the operators' tasks and providing input data on the composition of future SMART operators. Background: SMART is a nuclear reactor for export which needs operators who can satisfy both safety and economic feasibility. Therefore, this study is fundamental research on the composition of operators and this research analyzed SMART tasks in terms of human safety performance. Method: After analyzing 10 SMART EOG in hierarchical task analysis, this study classified task performance types according to task requirements of NUREG-0711 (Rev.3). Results: This study found the task frequency of SMART EOG and 12 operating task types. Conclusion: This study expects that human performance can be improved by analyzing the personal errors, which have the highest task frequency among 12 operating task types. Application: The results of this study can be applied as base data when licensing needs to be acquired.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1998년도 제13차 학술회의논문집
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• pp.295-300
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• 1998

In this paper, a real time diagnostic algorithm fur estimating the impact location by loose parts is proposed. It is composed of two modules such as the alarm discrimination module (ADM) and the impact-location estimation module(IEM). ADM decides whether the detected signal that triggers the alarm is the impact signal by loose parts or the noise signal. When the decision from ADM is concluded as the impact signal, the beginning time of burst-type signal, which the impact signal has usually such a form in time domain, provides the necessary data fur IEM. IEM by use of the arrival time method estimates the impact location of loose parts. The overall results of the estimated impact location are displayed on a computer monitor by the graphical mode and numerical data composed of the impact point, and thereby a plant operator can recognize easily the status of the impact event. This algorithm can perform the diagnosis process automatically and hence the operator's burden and the possible operator's error due to lack of expert knowledge of impact signals can be reduced remarkably. In order to validate the application of this method, the test experiment with a mock-up (flat board and reactor) system is performed. The experimental results show the efficiency of this algorithm even under high level noise and potential application to Loose Part Monitoring System (LPMS) for improving diagnosis capability in nuclear power plants.

• Corrosion Science and Technology
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• 제18권4호
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• pp.148-153
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• 2019

Pure nickel with a thickness of 1 mm was plated on type 304 stainless steels and low alloy steels (JIS G3131 SPHC) by electrolytic plating method in a circulating plating bath. Plating performance, mechanical properties, and surface characteristics were evaluated in terms of pretreatment process, anode material, pH, current density, and flow rate of the plating solution. Addition of hydrochloric acid during pre-treatment process improved the adhesion performance of plating. To improve plating efficiency, it is desirable to use S-nickel rather than electrolytic nickel. The use of S-nickel was also confirmed to be desirable for maintaining the pH and concentration of the plated solution. The defect of the plating using S-nickel anode produced pit on the surface. However, it is believed that proper control can be obtained by increasing the flow rate. Internal stress and hardness values of electrolytic nickel plating according to current density need to be carried out with further studies.

• Nuclear Engineering and Technology
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• 제5권1호
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• pp.55-64
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• 1973

압축변형시킨 순천(0.007% 탄소포함)을 8$0^{\circ}C$이하에서 저온열처리(ageing)한 후, 열처리시간에 따라 증가하는 저항복점(lower yield point)의 변화를 압축시험으로 측정하여 그 strain ageing 효과를 조사했다. 본 실험에서 압축변형된 순철의 strain ageing 속도는 저항복응력의 증가가 그치고 중가의 60% 정도에 이를 때까지 열처리시간의 2/3승에 비례했으나, 이미 알려진 인장변형의 경우에서보다 느렸다. 이 60%의 증가는 순철을 6$0^{\circ}C$에서 약 5시간 열처리함으로써 얻었다. 압축변형된 순철의 strain ageing을 위한 활성화에너지는 열처리의 초기단계(6$0^{\circ}C$에서 약 30분)에서 21,500 cal/mole이었는데, 이것은 인장변형의 경우에서 알려진 것보다 대략 10%가 큰 값이다. 이 증가는 잔유응력에 의해 결정내에 형성되는 strain field로 설명되었다. 열처리공정의 둘깨단계(6$0^{\circ}C$에서 약 5시간까지 계속되는)에서는 그 활성화에너지가 다소 감소되는 경향이 있었다.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.401-422
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• 2005

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.317-324
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• 1987

본 논문은 가압 경수형 원자로의 제어봉 이탈사고와 같이 공간 의존적 과도특성 해석에 필히 요구되는 가상적 사고 분석을 위한 핵적 동특성 코드의 개발을 위한 것이다. 본 논문에서는 1.5군 중성자 화산 방정식에 의거한 수정형 Borresen 모형을 핵적 동특성 모델로 잡고 이를 공간의존적 과도특성해석에 응용할 수 있도록 수식화 하여 고리 1호기 초기 노심의 가상적인 제어봉 이탈 사고해석에 응용했다. 본 사고 해석에 채택한 수정형 Borresen 모형에 대한 계산 정밀도의 검증을 위해 출력 분포 및 제어봉가등 계산결과를 고리 1호기 초기 노심의 노물리 실험자료와 비교했고 공간의존적 사고해석에 있어서 중시되는 핵적 동특성 방정식의 계산 효율성을 검토했다. 그리고 이 결과를 토대로 수정형 Borresen 모형이 제어봉 이탈사고, 증기관 파탄사고 등과 같은 공간의존적 사고해석에 유용하게 이용될 수 있다는 것을 보였다.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.121-128
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• 2009

국내 원전의 계획예방정비기간중에 원자로계통의 개방과정에서 원자로건물내 공기 중으로 누설된 $^{131}I$의 채내 흡입으로 원전 종사자의 내부피폭이 발생하였다. 이에 따라 원전에서 보유하고 있는 전신계측기(Whole Body Counter)를 이용하여 방사선작업 종료 후 즉시 원전종사자의 체내에 침적된 내부방사능을 측정하였고, 수일 경과 후 재측정하였다. 이러한 전신계측결과를 이용한 섭취량 산정 값을 원전종사자가 출입한 원자로 건물 내 공기 중의 $^{131}I$ 방사능 농도 측정결과와 원자로건물 출입기록에 근거하여 계산된 $^{131}I$ 채내 섭취량과 비교 평가하였다. 그 결과 전신계측기를 이용한 채내 방사능측정 결과와 공기중 농도를 이용한 섭취량 산정 결과는 비교적 잘 일치하는 것으로 평가되였다.