• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.573-578
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• 1997

기존 냉각수 주입 계통의 안전성 및 신뢰도를 높이려는 노력에서 미국, 일본, 이태리 등 여러 나라에서는 증기 주입기에 대한 분석 기법 개발과 실험을 통한 연구가 활발하게 진행되고 있다. 이들이 이상 유동 현상이 일어나는 혼합 노즐을 해석하기 위해 사용한 방법은 제어체적 분석기법으로, 이 분석 방법에서는 응축 메커니즘을 고려할 수 없다. 본 연구에서는 이러한 사항을 개선하기 위해 혼합 노즐 해석시 one dimensional two fluid model을 적용하여 상간의 운동량 및 에너지 전달을 고려함으로써 비교적 정확하게 해석 모델 및 코드를 개발하였고, 출력 변수인 방출 압력과 방출 온도를 계산하였으며, 개발한 해석 코드를 검증 및 분석을 하기 위해 ANL 실험자료를 근거로 하였다. 개발한 코드에 의해 계산된 방출 압력은 ALN 실험치에 비해 높은 경향을 보여 주었지만 ANL에서 제어 체적 분석 기법을 적용하여 이론식으로 계산한 값보다는 비교적 정확한 경향을 보여 주었다.

• Proceedings of the Korea Information Processing Society Conference
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• 2000.10b
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• pp.903-906
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• 2000

웨이브렛 영역에서 벡터 양자화(vector quantization)를 수행하여 생성된 VQ 인덱스들을 엔트로피 부호화(entropy coding)하면 영상의 코딩 효율을 높일 수 있다. 본 논문에서는 벡터 양자화 이전에 VQ 인덱스들의 중복성을 높이기 위해 다중해상도 코드북의 코드 워드들을 에너지 크기 순으로 재정렬하는 기법을 제안한다. 코드 워드들의 평균과 편차를 이용한 재정렬 방법과 제안된 기법을 벡터 양자화 후 생성되는 VQ 인덱스에 DPCM/Huffman 기법을 적용하여 각각에 대한 코딩 효율을 비교한다.

• The Journal of the Acoustical Society of Korea
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• v.21 no.2
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• pp.167-172
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• 2002

The EVRC (Enhanced Variable Rate Codec) is currently in service as a speech cosec in digital cellular systems in North America and Korea. In the EVRC, the LSP (Line Spectral Pairs) related to energy distribution of speech signals in the frequency domain are coded by weighted split vector quantization. Considering that the LSP codebooks might be trained with the language of the develop country of the codebooks or English, it is expected that codebooks trained with Korean provide the performance improvements in the communication in Korean. In this paper, the EVRC LSP codebooks are designed with korean adopting the LBG algorithm based vector quantization, and the performance improvement of the vector quantization and the accompanying speech quality improvement are demonstrated by spectral distortion, SNR and SegSNR measurements, respectively.

• Proceedings of the KIEE Conference
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• 2000.11a
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• pp.262-264
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• 2000

원자로 내에서 연소 중인 핵연료나 저장 또는 재처리 중인 사용후핵연료의 성분으로서 시설의 공정설계, 안전성분석 및 차폐설계에 중요한 입력자료가 되는 핵분열생성물질, 방사화생성물 및 악티나이드의 핵종 농도와 이에 대응하는 방사능 강도의 기기 별 시간변 화율을 해석할 수 있는 코드 개발할 목적으로 MULTISAMS 정상 평형상태 모델을 구현하였다. MULTISAMS 코드의 반응공정 모델은 서로 연결되어 있으며 내부에 방사성물질의 혼합유체가 순환하는 세 종류의 반응기(원자로, 열교환기 및 화학반응기) 계통에서 자연적 또는 설계에 의해 일어나는 현상으로서; 반응기 간의 물질 흐름; 각 반응기 내에서 방사성 붕괴, 변환, 이동과 중성자 흡수 및 핵분열; 외부로부터 특정 핵종의 유입혹은 유출을 고려한 시간종속 핵종농도보존방정식 이론에 근거한다. 코드의 유용성 및 신뢰성을 검증하기 위해 현재 개념설계가 진행 중인 AMBIDEXTER원자력 에너지시스템을 대상으로 ORIGEN2 계산과 비교하였다. 두 코드 간의 입력조건과 배경이론차이점 때문에 절대적 비교가 불가능하므로 단순이론의 중간매개코드로서 SAMS를 이용한 2단계 비교방법을 따랐다. 결론은 MULTISAMS는 ORIGEN2 계산의 수렴치와 근사하게 일치하면서 ORIGEN2 가 다룰 수 없는 핵주기 연속후처리공정의 정상가동 시 핵종 평형농도를 기기 별로 계산할 수 있다는 장점을 확인하였다.

• Journal of the Institute of Electronics and Information Engineers
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• v.52 no.6
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• pp.3-9
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• 2015

We investigate rate scheduling and power allocation problem for a delay constrained CDMA systems. Specifically, we determine an energy efficient scheduling policy, while each user maintains the short term (n time slots) average throughput. We consider a multirate CDMA system where multirate is achieved by multiple codes. Each code can be interpreted as a virtual user. The aim is to schedule the virtual users into each time slot, such that the sum of transmit energy in n time slots is minimized. We then show that the total energy minimization problem can be solved by a shortest path algorithm. We compare the performance of the optimum scheduling with that of TDMA-type scheduling.

• Explosives and Blasting
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• v.37 no.3
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• pp.25-33
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• 2019

Controlling the blast-induced damage zone(BDZ) in mining excavation is a significant issue for the safety of employees and the maintenance of facilities. Numerous studies have been conducted to accurately predict the BDZ in underground mining. This study employed the dynamic fracture process analysis (DFPA) to estimate the BDZ from a single hole blasting. The estimated BDZ were compared with the results obtained by Swedish empirical equation. The DFPA was also used to investigate the control mechanism of BDZ and fracture plane formation around perimeter holes for underground mining blasting.

• Journal of Energy Engineering
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• v.10 no.4
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• pp.299-309
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• 2001

This shell-and-tube heat exchanger design code based on stream analysis method was developed to design accurate and advanced heat exchangers. Several geometry factors which affect the heat exchanger design was explained. Stream analysis method to calculate flow fraction of each stream and heat exchanger design flow chart was introduced. Performance of developed simulation code was compared with Delaware09, Delaware10, DongHwa and ANL experimental data. The statistical results of performance evaluation indicated that most data points are predicted within $\pm$30%. But the pressure loss was over predicted.

• Journal of Energy Engineering
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• v.24 no.4
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• pp.175-182
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• 2015

A methodology to determine the most conservative initial condition based on random sampling of operation parameters is established, in which a best-estimate computer code is adopted to minimize the conservatism in code models. To validate the applicability of the suggested method, safety evaluation for a transient of loss of condenser vacuum in a pressurized water reactor is performed. One-hundred different initial conditions are generated by MOSAIQUE program automatically and the peak pressure for the most conservative case is determined from transient analyses. The safety margin obtained with the new approach is almost equivalent to the values determined with the existing methodologies. It is found that the time and human resources required for the safety evaluation could be reduced with the suggested approach.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.198-204
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• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.95-105
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• 2014

The CUPID code has been developed for a transient, three-dimensional, two-phase flow analysis at a component scale. It has been validated against a wide range of two-phase flow experiments. Especially, to assess its applicability to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor, it was validated using the experimental data of the 1/4-scaled facility of a Calandria vessel at the STERN laboratory. In this study, a preliminary thermal-hydraulic analysis of the CANDU reactor moderator tank using the CUPID code is carried out, which is based on the results of the previous studies. The complicated internal structure of the Calandria vessel and the inlet nozzle was modeled in a simplified manner by using a porous media approach. One of the most important factors in the analysis was found to be the modeling of the tank inlet nozzle. A calculation with a simple inlet nozzle modeling resulted in thermal stratification by buoyance, leading to a boiling from the top of the Calandria tank. This is not realistic at all and may occur due to the lack of inlet flow momentum. To improve this, a new nozzle modeling was used, which can preserve both mass flow and momentum flow at the inlet nozzle. This resulted in a realistic temperature distribution in the tank. In conclusion, it was shown that the CUPID code is applicable to thermal-hydraulic analysis of the CANDU reactor moderator tank using the cost-effective porous media approach and that the inlet nozzle modeling is very important for the flow analysis in the tank.