• The Transactions of the Korea Information Processing Society
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• v.5 no.12
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• pp.3127-3137
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• 1998

COBRA 표준명세는 표준을 만족하는 구현에서 제공해야 할 기능뿐만 아니라 서비스 제공 모듈의 사용자 인터페이스도 IDL을 사용하여 엄격하게 정의하고 있다. CORBA 표준에 대한 확신과 신뢰성을 가지기 위해서는 IDL(Interface Definition Language)로 기술된 표준명세를 정형화하고 수학적으로 엄격히 증명할 필요가 있다. 본 논문에서는 CORBA 표준을 정형적으로 명세하고 검증할 방법을 제시한다. 먼저 표준모듈을 Larch/CORBA IDL(LCB)를 사용하여 정형적으로 명세하고, LCB의 의미론에 준하여 LCB 명세를 LSL(Larch Shared language)로 변환한다. 변환한 LCB 명세와 LSL 증명논리를 사용하여 특성을 수학적으로 증명한다. 변환기반의 LCB 의미론을 정립하여 제안한 방법의 이론적 바탕을 마련하고 CORBA 이름서비스명세에 실제 적용하여 그 효용성을 보인다.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.596-597
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• 2005

• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Journal of Intelligence and Information Systems
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• v.3 no.1
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• pp.143-160
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• 1997

현재 우리 군에서는 첨단과학무기를 이용한 전투력을 신속히 집중, 전환시키고 효과적으로 통합 운용해야하는 각급 제대의 지휘관 및 참모의 지휘통제능력 향상을 위하여 첨단 컴퓨터장비를 이용하여 시뮬레이션 기법을 통한 워게임 모델을 개발하여 이를 이용한 훈련을 실시하고 있다. 이 워게임 모델중 지상전투의 가장 기본이 되는 근접전투 시뮬레이션은 미국에서 개발도입된 "COBRA" 시스템을 이용하고 있으나 한국실정에 맞는 시스템으로 확장 및 유지보수가 어렵고, 상위시스템의 서브시스템으로만 운영되고있어 자체 교육훈련 및 전투분석을 위한 단독시스템으로 운영이 어려운 실정이다. 본 논문에서는 이러한 문제점을 극복하고, 방대한 양의 지식을 효율적이고 효과적으로 표현할 수 있으며 시스템의 확장 및 유지보수가 용이하고 우리실정에 적합한 전투 훈련을 실시하도록 지원하는 워게임(근접전투) 지원용 멀티미디어 전문가시스템을 개발하였다. 본 논문에서 개발한 전문가시스템은 쌍방이 부대들의 근접전투를 실시할 때 실전에서 나타날 수 있는 가능한 모든 상황의 데이터를 이용하여 전투상황을 분석하며, 기존의 획일적이고 단순한 형태로 결과를 판정하던 것을 전투원의 사기, 체력, 전투한계 등 심리적 요소까지 고려함으로써 새로이 변화되는 전쟁양상에 쉽게 적응할 수 있는 확장성 및 유지보수가 용이하며 시스템 단독으로 운영하여 반복적으로 전투를 분석하고 교육훈련을 실시하도록 함으로써 실전적이고 실질적인 근접전투 워게임지원이 가능하다. 본 논문에서는 전문가 시스템을 개발함에 있어서 지식베이스 모듈, 추론엔진 모듈 및 설명 모듈은 전문가 시스템 개발도구인 Smart Elements를 이용하여 구축하였으며, 사용자 인터페이스 모듈은 멀티미디어 저적도구인 툴북 3.0을 이용하였으며, 마지막으로 전체적인 모듈은 API를 이용 통합하여 하나의 응용소프트웨어를 생성하였다.

• Journal of the Korean Society of Clothing and Textiles
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• v.41 no.3
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• pp.537-547
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• 2017

This study evaluates the wearing performance of a self-contained breathing apparatus (SCBA) belt for firefighters in Korea. A SCBA belt design was suggested based on the wearing evaluation of a SCBA belt; subsequently, prototypes I and II were created. A wearing evaluation of prototypes with improved design and design preference was performed. Six designs elements of the SCBA belt for firefighters were suggested based on the survey results of wearing the SCBA belt and for the SCBA belt design preference for use by firefighters. First, belt material should be made of black high-strength aramid textiles. In addition, Velcro should be used to attach and detach retroreflective and fluorescent materials along with various colors for visibility. Second, the chest belt should be made of the same material used for other parts; in addition, the chest belt should be moved to the center for center of gravity and a cobra buckle should be applied. Third, an O-ring should be applied to the back and the belt connected to the O-ring should distribute the weight in six axes. Fourth, a detachable air respirator should be able to separate by using upper and lower cobra buckles. Fifth, a separable leg belt and a detachable pocket are also suggested. Sixth, a ring for walkie-talkies, alarms and equipment as a fabric ring are also suggested. Prototype III with an improved design was created based on the results of the design suggestion.

• Clinical and Experimental Pediatrics
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• v.60 no.11
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• pp.353-358
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• 2017

Purpose: To develop and evaluate a simple screening tool to assess hearing loss in newborns. A derived score was compared with the standard clinical practice tool. Methods: This cohort study was designed to screen the hearing of newborns using transiently evoked otoacoustic emission and auditory brain stem response, and to determine the risk factors associated with hearing loss of newborns in 3 tertiary hospitals in Northern Thailand. Data were prospectively collected from November 1, 2010 to May 31, 2012. To develop the risk score, clinical-risk indicators were measured by Poisson risk regression. The regression coefficients were transformed into item scores dividing each regression-coefficient with the smallest coefficient in the model, rounding the number to its nearest integer, and adding up to a total score. Results: Five clinical risk factors (Craniofacial anomaly, Ototoxicity, Birth weight, family history [Relative] of congenital sensorineural hearing loss, and Apgar score) were included in our COBRA score. The screening tool detected, by area under the receiver operating characteristic curve, more than 80% of existing hearing loss. The positive-likelihood ratio of hearing loss in patients with scores of 4, 6, and 8 were 25.21 (95% confidence interval [CI], 14.69-43.26), 58.52 (95% CI, 36.26-94.44), and 51.56 (95% CI, 33.74-78.82), respectively. This result was similar to the standard tool (The Joint Committee on Infant Hearing) of 26.72 (95% CI, 20.59-34.66). Conclusion: A simple screening tool of five predictors provides good prediction indices for newborn hearing loss, which may motivate parents to bring children for further appropriate testing and investigations.

• Journal of the Korean Association of Oral and Maxillofacial Surgeons
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• v.31 no.6
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• pp.468-473
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• 2005

Head and neck squamous cell carcinoma (HNSCC) is the sixth most common malignancy worldwide. The molecular mechanisms involved in the development and progression of these carcinomas are not well known. Abnormalities of genomic methylation patterns have been attributed a role in carcinogenesis and local de novo methylation at tumor suppressor loci was held to be involved in silencing of tumor suppressor genes. Using Ms APPCR, we previously isolated a hypermethylated fragment corresponded to the 5' end of TPEF gene from primary liver and lung cancer cells. To confirm the inactivation of TPEF gene by hypermethylation in HNSCC, we investigated correlation between methylation pattern and expression of TPEF in 10 HNSCC cell lines. In methylation analysis such as combined-bisulfite restriction analysis(COBRA) and bisulfite sequencing, only RPMI 2650 showed none methylated pattern and another 9 cell lines showed dense methylation. The TPEF gene expression level analysis using RT-PCR showed that these 9 cell lines had not or significantly low expression levels of TPEF as compared with RPMI 2650. In addition, the increase of TPEF reexpression by 5-AzaC as demethylating agent in 9 cell lines also indicated that TPEF expression was regulated by hypermethylation. These results of this study demonstrate that epigenetic silencing of TPEF gene by aberrant methylation could play an important role in HNSCC carcinogenesis.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.256-266
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• 1985

The loss of coolant flow accident is analyzed for the pressurized water reactor of Korea Nuclear Unit-1. The loss of coolant flow accident is classified into three types in accordance with its severity; partial loss of coolant flow, complete loss of coolant flow and pump locked rotor accident. Analysis has been carried out in three stages; system transient and average core analysis, DNBR calculation and hot spot analysis. The purpose of developing KTRAN is to simulate the transient fast. For the DNBR calculation, the thermal hydraulic codes, SCAN and COBRA IV-1, are adopted. And for the hot spot analysis, the fuel thermal transient code LTRAN is employed. This code system should be fast responding to the transient analysis. In case the transient occurs, severity comes within a couple of seconds. So response should be fast to accomodate the following sequence of the accident. Unfortunately this purpose could not be achieved by KTRAN. However, the calculated results are well comparable with FSAR results in range. Thereby, the effectiveness of KTRAN code analysis in this type of accident is proven.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.155-160
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• 1984

Analyzed is the thermal margin of the Korea Nuclear Unit 1 (KNU-1) reactor core consisting of either 14 x 14 standard fuel assemblies (SFA) or optimized fuel assemblies (OFA). Employed for the analysis are two different thermal design methods; traditional and statistical thermal design method. Compared to the traditional design thermal method, the statistical thermal design method improves the core thermal margin utilizing best-estimate values for the core operating parameters combining their uncertainties in a statistical manner. Calculations are performed using a steady state and transient thermal-hydraulic analysis computer program, COBRA-IV-i. Calculated results show that the statistical thermal design method significantly improves the thermal margin and satisfies the core thermal design base of the KNU-1 SFA and OFA core. However, the thermal design base can not be met, if the traditional thermal design method is employed for the OFA role analysis.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.66-74
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• 1990

The thermal-hydraulic characteristics of the 17$\times$17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17$\times$17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.