• 대한전자공학회논문지SD
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• 제41권3호
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• pp.83-92
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• 2004

본 논문에서는 새로운 interpolation-2 방식의 비교기 구조를 제안하여 칩 면적과 전력 소모를 줄이며 오류정정 회로를 내장하는 6-비트 70㎒ ADC를 설계하였다. Interpolation 비교기를 적용하지 않은 flash ADC의 경우 2n개의 저항과 2n -1개의 비교기가 사용되며 이는 저항의 수와 비교기의 수에 비례하여 많은 전력과 큰 면적을 필요로 하고 있다. 또한, interpolation-4 비교기를 적용한 flash ADC는 면적은 작으나 단조도, SNR, INL, DNL 특성이 떨어진다는 단점이 있었다. 본 논문에서 설계한 interpolation-2 방식의 ADC는 저항, 비교기, 앰프, 래치, 오류정정 회로, 온도계코드 디텍터와 인코더로 구성되며, 32개의 저항과 31개의 비교기를 사용하였다. 제안된 회로는 0.18㎛ CMOS 공정으로 제작되어 3.3V에서 40mW의 전력소모로 interpolation 비교기를 적용하지 않은 flash ADC에 비해 50% 개선되었으며, 칩 면적도 20% 감소되었다. 또한 노이즈에 강한 오류정정 회로가 사용되어 interpolation-4 비교기를 적용한 flash ADC 에 비해 SNR이 75% 개선된 결과를 얻었다.

• Journal of Radiation Protection and Research
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• 제28권2호
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• pp.79-85
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• 2003

한국원자력연구소에서 개발한 $CaSO_4:Dy,P$를 이용하여 ICRU에서 권고하는 Hp(10)을 평가하고자 하였다. $CaSO_4:Dy,P$는 광자에 내하여 높은 에너지 의존성을 지니므로 Hp(10) 평가를 위해서는 필터물질을 이용하여 모든 에너지에 대하여 반응도를 적절히 보상하여야 한다. ISO에서 권고하는 열형광선량계의 성능요건을 만족하는 필터설계를 위해 몬테칼로 코드를 이용한 이론적 계산과 검증 실험을 수행하였다. 완성된 필터 하에서 $CaSO_4:Dy,P$ 소자는 $20{\sim}662keV$에 이르는 광자 에너지 범위에서 $0.75{\sim}1.0$의 상대반응도를 보였다. 특히 직경이 큰 후방필터를 설계함으로써 전방필터의 두께를 감소하고 필터 조합을 단순화하였으며 테이퍼를 도입하여 선량계의 방향의존성을 대폭 개선할 수 있었다.

• Journal of Magnetics
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• 제21권2호
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• pp.183-186
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• 2016

In this study, we report on an investigation of proton radiation resistance of 410 martensitic stainless steels under 3 MeV proton with the doses ranging from $1.0{\times}10^{15}$ to $1.0{\times}10^{17}p/cm^2$ at the temperature 623 K. Vibrating sample magnetometer (VSM) and X-ray diffractometer (XRD) were used to study the variation of magnetic properties and structural damages by virtue of proton irradiation, respectively. VSM and XRD analysis revealed that the 410 martensitic stainless steels showed proton radiation resistance up to $10^{17}p/cm^2$. Proton energy degradation and flux attenuations in 410 stainless steels as a function of penetration depth were calculated by using Stopping and Range of Ions in Matter (SRIM) code. It suggested that the 410 stainless steels have the radiation resistance up to $5.2{\times}10^{-3}$ dpa which corresponds to neutron irradiation of $3.5{\times}10^{18}n/cm^2$. These results could be used to predict the maintenance period of SUS410 stainless steels in fission power plants.

• International Journal of Fluid Machinery and Systems
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• 제3권1호
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• pp.11-19
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• 2010

The performance of the crossover system of a centrifugal compressor stage consisting of static components of $180^{\circ}$ U-bend, return channel vanes and exit ducting with a $90^{\circ}$ bend is investigated. This study is confined to the assessment of performance of the crossover system by varying the shape of the return channel vanes. For this purpose two different types of Return Channel Vanes (RCV1 and RCV2) were experimentally investigated. The performance of the crossover system is discussed in terms of total pressure loss coefficient, static pressure recovery coefficient and vane surface pressure distribution. The experimentation was carried out on a test setup in which static swirl vanes were used to simulate the flow at the exit of an actual centrifugal compressor impeller with a design flow coefficient of 0.053. The swirl vanes are connected to a mechanism with which the flow angle at the inlet of U-bend could be altered. The measurements were taken at five different operating conditions varying from 70% to 120% of design flow rate. On an overall assessment RCV1 is found to give better performance in comparison to RCV2 for different U-bend inlet flow angles. The performance of RCV2 was verified using numerical studies with the help of a CFD Code. Three dimensional sector models were used for simulating the flow through the crossover system. The turbulence was predicted with standard k-$\varepsilon$, 2-equation model. The iso-Mach contour plots on different planes and development of secondary flows were visualized through this study.

• 한국산학기술학회논문지
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• 제11권3호
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• pp.920-925
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• 2010

본 논문은 유무선 네트워크와 댁내 가전기기로 구성된 홈 네트워크 시스템은 각종 외부 위협요소로부터의 사이버 공격 대상이 될 수 있을 뿐만 아니라 해킹, 악성코드, 웜 바이러스, DoS공격, 통신망 도청 등의 보안취약성을 가지고 있다. 이로 인해 사생활 침해, 개인정보의 노출, 개인정보의 도용 등 많은 문제가 발생한다. 따라서 홈 네트워크상에서 해당 사용자의 자산 및 개인정보를 보호할 수 있는 보안 프로토콜의 필요성은 점차 증대되어가고 있다. 따라서 본 논문에서는 공개키 인증서를 이용하여 사용자를 인증하고 인증된 정보를 기반으로 해당 기기에 대한 권한을 차등 부여함으로서 허가 받지 않은 사용자로부터의 댁내 자산 및 개인 정보를 보호할 수 있는 사용자 인증과 접근제어 기술을 이용한 홈 네트워크 보안 프로토콜을 설계하고 제안한다.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.867-874
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• 2009

Solubility of a radionuclide is important for defining the release source term of a radioactive waste in the safety and performance assessments of a radioactive waste repository. When the pH and redox potential of the KURT groundwater were changed by an electrical method, the concentrations of uranium and thorium released from $UO_2$(cr) and $ThO_2$(cr) at alkali pH(8.1 ${\sim}$ 11.4) and reducing potential (Eh < -0.2 V) conditions were less than $10^{-7}mole/L$. Unexpectedly, the concentration of tetravalent thorium is slightly higher than that of uranium at pH = 8.1 and Eh= -0.2 V conditions, and this difference may be due to the formation of hydroxide-carbonate complex ions. When $UO_2$(s) and $UO_2$(am, hyd.), and $ThO_2$(s) and $Th(OH)_4(am)$ were assumed as solubility limiting solid phases, the concentrations of uranium and thorium in the KURT groundwater calculated by the PHREEQC code were comparable to the experimental results. The dominating aqueous species of uranium and thorium were presumed as $UO_2(CO_3)_3^{4-}$ and $Th(OH)_3CO_3^-$ at pH = 8.1 ${\sim}$ 9.8, and $UO_2(OH)_3^-$ and $Th(OH)_4(aq)$ at pH = 11.4.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• 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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• 제53권9호
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• pp.3051-3057
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• 2021

The concrete is considered as an important radiation shielding material employed widely in nuclear reactors, particle accelerators, laboratory hot cells and other different radiation sources. The present research is dedicated to the shielding properties study of the ordinary concrete reinforced with different weight fractions of lead oxide micro/nano particles. Lead oxide particles were fabricated by chemical synthesis method and their properties including the average size, morphological structure, functional groups and thermal properties were characterized by XRD, FESEM-EDS, FTIR and TGA analysis. The gamma ray mass attenuation coefficient of concrete composites has been calculated and measured by means of the Monte Carlo simulation and experimental methods. The simulation process was based on the use of MCNP Monte Carlo code where the mass attenuation coefficient (μ/ρ) has been calculated as a function of different particle sizes and filler weight fractions. The simulation results showed that the employment of the lead oxide filler particles enhances the mass attenuation coefficient of the ordinary concrete, drastically. On the other hand, there are approximately no differences between micro and nano sized particles. The mass attenuation coefficient was increased by increasing the weight fraction of nanoparticles. However, a semi-saturation effect was observed at concentrations more than 10 wt%. The experimental process was based on the fabrication of concrete slabs filled by different weight fractions of nano lead oxide particles. The mass attenuation coefficients of these slabs were determined at different gamma ray energies using ²²Na, ¹³⁷Cs and ⁶⁰Co sources and NaI (Tl) scintillation detector. The experimental results showed that the HVL parameter of the ordinary concrete reinforced with 5 wt% of nano PbO particles was reduced by 64% at 511 keV and 48% at 1332 keV. Reasonable agreement was obtained between simulation and experimental results and showed that the employment of nano PbO particles is more efficient at low gamma energies up to 1Mev. The proposed concrete is less toxic and could be prepared in block form instead of toxic lead blocks.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.949-959
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• 2021

The current study aims to explore the shielding properties of multi-component borate-based glass series. Seven glass-samples with composition of (80-y)H₃BO₃-10ZnO-10Na₂O-yBaO where (y = 0, 5, 10, 15, 20, 25 and 30 mol.%) were synthesized by melt-quench method. Various shielding features for photons, neutrons, and protons were determined for all prepared samples. XCOM, Phy-X program, and SRIM code were performed to determine and explain several shielding properties such as equivalent atomic number, exposure build-up factor, specific gamma-ray constants, effective removal cross-section (Σ_R), neutron scattering and absorption, Mass Stopping Power (MSP) and projected range. The energy ranges for photons and protons were 0.015-15 MeV and 0.01-10 MeV, respectively. The mass attenuation coefficient (μ/ρ) was also determined experimentally by utilizing two radioactive sources (¹⁶⁶Ho and ¹³⁷Cs). Consistent results were obtained between experimental and XCOM values in determining μ/ρ of the new glasses. The addition of BaO to the glass matrix led to enhance the μ/ρ and specific gamma-ray constants of glasses. Whereas the remarkable reductions in Σ_R, MSP, and projected range values were reported with increasing BaO concentrations. The acquired results nominate the use of these glasses in different radiation shielding purposes.