• Geomechanics and Engineering
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• 제16권1호
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• pp.21-34
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• 2018

Discontinuities considerably affect the mechanical and hydraulic properties of rock mass. These properties of the rock mass are influenced by the geometry of the discontinuities to a great extent. This paper aims to render an account of the geometrical parameters of several discontinuity sets related to the surrounding rock mass of Rudbar Lorestan Pumped Storage Power Plant powerhouse cavern making use of the linear and areal (circular and rectangular) sampling methods. Taking into consideration quite a large quantity of scanline and the window samplings used in this research, it was realized that the areal sampling methods are more time consuming and cost-effective than the linear methods. Having corrected the biases of the geometrical properties of the discontinuities, density (areal and volumetric) as well as the linear, areal and volumetric intensity accompanied by the other properties related to four sets of discontinuities were computed. There is an acceptable difference among the mean trace lengths measured using two linear and areal methods for the two joint sets. A 3D discrete fracture network generation code (3DFAM) has been developed to model the fracture network based on the mapped data. The code has been validated on the basis of numerous geometrical characteristics computed by use of the linear, areal sampling methods and volumetric method. Results of the linear sampling method have significant variations. So, the areal and volumetric methods are more efficient than the linear method and they are more appropriate for validation of 3D DFN (Discrete Fracture Network) codes.

• Journal of Cancer Prevention
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• 제23권4호
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• pp.183-190
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• 2018

Background: As the number of big-cohort studies increases, validation becomes increasingly more important. We aimed to validate administrative database categorized as colorectal cancer (CRC) by the International Classification of Disease (ICD) 10th code. Methods: Big-cohort was collected from Clinical Data Warehouse using ICD 10th codes from May 1, 2003 to November 30, 2016 at Seoul National University Bundang Hospital. The patients in the study group had been diagnosed with cancer and were recorded in the ICD 10th code of CRC by the National Health Insurance Service. Subjects with codes of inflammatory bowel disease or tuberculosis colitis were selected for the control group. For the accuracy of registered CRC codes (C18-21), the chart, imaging results, and pathologic findings were examined by two reviewers. Sensitivity, specificity, positive predictive value (PPV), and negative predictive value (NPV) for CRC were calculated. Results: A total of 6,780 subjects with CRC and 1,899 control subjects were enrolled. Of these patients, 22 subjects did not have evidence of CRC by colonoscopy, computed tomography, magnetic resonance imaging, or positron emission tomography. The sensitivity and specificity of hospitalization data for identifying CRC were 100.00% and 98.86%, respectively. PPV and NPV were 99.68% and 100.00%, respectively. Conclusions: The big-cohort database using the ICD 10th code for CRC appears to be accurate.

• 한국항공운항학회지
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• 제16권4호
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• pp.12-19
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• 2008

GIOVE-B is the second experimental Galileo satellite for the Galileo In-Orbit Validation, which was launched by a Soyuz/Fregat rocket departing from the Baikonur cosmodrome in Kazakhstan on 27 April and now operated successfully. This paper presents the results obtained from processing of the E5a signal transmitted from the GIOVE-B. The acquisition and tracking of the data and pilot channels are performed by the E5a software receiver implemented by ETRI. Moreover, the paper suggests the GIOVE-B E5a-I/E5a-Q secondary code, which is determined by analyzing the correlation output of the primary correlator using the primary code.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3801-3807
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• 2023

Outcomes of the project "Comprehensive Verification of the FALCON Code for Calculation of Nuclear Fuel Temperature" relating to calculation of fuel temperature during relatively fast thermal transients are presented. Good prediction capabilities of the FALCON MOD01 code coupled with the GRSW-A code are shown as applied to the data of the TF3 and TF5 tests from the Transient Temperature Experiment IFA-507. The IFA-507 related dataset of the OECD/NEA International Fuel Performance Experiments (IFPE) Database is extended by the reconstructed dynamics of the axial power distribution in the rods during the transient phase of the experiment. Based on the code calculation, the time constant of the thermal fuel response to a power transient is estimated.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• 정보처리학회논문지C
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• 제13C권4호
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• pp.397-404
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• 2006

소프트웨어 공격이 성공하기 위해서는 공격 코드가 프로그램의 주소 공간에 주입된 후 프로그램의 제어 흐름이 공격 코드 위치로 변경되어야 한다. 프로그램의 주소 공간 중 코드 영역은 실행 중에 변경이 불가능하므로 공격 코드가 주입될 수 있는 곳은 데이터 영역 밖에 없다. 따라서 데이터 영역으로 프로그램의 제어가 변경될 경우 주입된 공격 코드로 제어가 옮겨 가는 공격이 발생한 것으로 판단할 수 있다. 따라서 본 논문에서는 프로그램의 제어 흐름과 관련된 CALL, JMP, RET 명령어의 목적 주소를 검사하여 제어가 옮겨갈 목적 주소가 프로그램의 실행 코드가 저장된 텍스트 영역이 아닌 데이터 영역일 경우 소프트웨어 공격이 발생한 것으로 간주하는 소프트웨어 공격 탐지 기법을 제안하였다. 제안된 방법을 이용하면 함수의 복귀주소뿐만 아니라 함수포인터, longjmp() 버퍼 등 프로그램 제어 흐름과 관련된 모든 데이터가 변경되었는지 점검할 수 있었기 때문에 기존 기법들보다 더 많은 종류의 공격을 탐지할 수 있었다.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• 한국항공우주학회지
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• 제35권3호
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• pp.183-195
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• 2007

이차원 범프 유동에 대한 다양한 예조건화 행렬의 수렴 특성을 살펴 Choi 와 Merkle 의 예조건화 행렬을 선택하여, 압축성 및 예조건화 Roe의 Riemann 해법의 수치 소산항을 수학적으로 비교하였다. 이 결과 코드의 구조는 동일하게 유지한 채, 고유치의 작은 수정만으로 압축성 해법을 예조건화 해법으로 이전할 수 있는 방법을 알 수 있었다. 아울러 점성 유동 영역에서의 안정성 및 정확성을 향상시키기 위하여 von Neumann 안정 조건 및 점성 자코비안을 고려하였으며, 개발된 코드는 표준 검증 문제에 적용하여 검증을 수행하였다.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.587-594
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• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.