• 한국정보통신학회:학술대회논문집
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• 한국정보통신학회 2012년도 춘계학술대회
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• pp.778-781
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• 2012

전자소자는 방사선 피폭으로 손상이 발생하므로 인공위성이나 우주선에 적용하기 위해서는 방사선 내성을 가진 내방사선 전자소자의 사용이 필수적이다. 국내에서 처음 수행한 과도방사선 시험평가 자료를 데이터베이스(DB)로 구축하였다. 이 DB는 웹기반으로 자료검색과 갱신이 가능하도록 설계되었고, 기존 외국 (NASA, ESA)의 공개 자료를 포함하여 총 695종의 데이터가 입력되어 있다. 기존 외국의 DB보다 검색이 효율적이고 실용적인 DB이다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.386-391
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• 1997

It is important to understand correctly a CHF under low flow condition for the purpose of enhancing the reactor safety and performance in the LWRs. The CHF experiments have been carried out for an internally heated vertical annulus in RCS loop facility. The experimental conditions cover ranges of pressure from 1.82 to 12.08 MPa, mass flux from 300 to 550kg/$m^2$. s and inlet subcooling of 210kJ/kg. The CHF data decrease with increasing pressure at high value of mass flux. For mass flux of about 300kg/$m^2$. s, the CHF rue little influenced by pressure. The CHF data are correlated well by using the dimensionless heat flux and dimensionless mass flux for a fixed inlet subcooling except the data group of 12.08 MPa. It seems that the Doerffer correlation and Katto correlation overestimate the CHF for low pressure and lower value of mass flux within this experimental ranges. The Bowling correlation gives a better prediction than the other two correlations.

• 환경영향평가
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• 제12권4호
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• pp.271-279
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• 2003

Temperature and humidity forecasting have been performed using artificial neural networks model(ANN). We composed ANN with multi-layer perceptron which is 2 input layers, 2 hidden layers and 1 output layer. Back propagation algorithm was used to train the ANN. 6 nodes and 12 nodes in the middle layers were appropriate to the temperature model for training. And 9 nodes and 6 nodes were also appropriate to the humidity model respectively. 90% of the all data was used learning set, and the extra 10% was used to model verification. In the case of temperature, average temperature before 15 minute and humidity at present constituted input layer, and temperature at present constituted out-layer and humidity model was vice versa. The sensitivity analysis revealed that previous value data contributed to forecasting target value than the other variable. Temperature was pseudo-linearly related to the previous 15 minute average value. We confirmed that ANN with multi-layer perceptron could support pollutant dispersion model by computing meterological data at real time.

• Nuclear Engineering and Technology
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• 제38권8호
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• pp.809-818
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• 2006

The existing experimental data related to the turbulent mixing factor in rod arrays is examined and a new definition of the turbulent mixing factor is introduced to take into account the turbulent mixing of fluids with various Prandtl numbers. The new definition of the mixing factor is based on the eddy diffusivity of energy. With this definition of the mixing factor, it was found that the geometrical parameter, ${\delta}_{ij}/D_h$ correlates the turbulent mixing data better than Sid, which has been used frequently in existing correlations. Based on the experimental data for a highly turbulent condition in square rod arrays, a correlation describing turbulent mixing dependent on the parameter ${\delta}_{ij}/D_h$ has been developed. The correlation is insensitive to the Re number and it takes into account the effect of the turbulent Prandtl number. The proposed correlation predicts a reasonable mixing even at a lower S/d ratio.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.899-906
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• 2018

The irradiation performance of $U_3Si$ dispersion fuel in an Al matrix, $U_3Si-Al$, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized $U_3Si$ powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4042-4051
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• 2021

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

• 비파괴검사학회지
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• 제26권4호
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• pp.246-254
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• 2006

방사선 측정분야에서 사용되는 감마 서베이미터의 교정데이터 자동처리를 위한 전산 프로그램을 개발하였다. 전산 프로그램은 Visual Basic을 기반으로 개발되었으며, 교정과정에 따라 단계별로 윈도우를 제작하고 코드화하였다. 교정 데이터의 자동처리를 위해 Microsoft Excel 프로그램을 제어하여 미리 자동 연산된 엑셀 셀 내에 데이터가 입력되도록 하였다. 개발프로그램 성능평가의 일환으로 검증된 데이터와 프로그램에서 출력된 데이터를 비교한 결과, 교정인자 산출 및 불확도 평가에서 동일한 결과가 나왔다. 또한, 개발프로그램을 교정업무에 적용시킨 결과, 업무의 효율성 및 정확성은 증가하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.896-908
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• 2022

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. In this regard, it is vital to provide credible HEPs based on firm technical underpinnings including (but not limited to): (1) how to collect HRA data from available sources of information, and (2) how to inform HRA practitioners with the collected HRA data. Because of these necessities, the U.S. Nuclear Regulatory Commission and the Korea Atomic Energy Research Institute independently developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. These systems provide unique frameworks that can be used to secure HRA data from full-scope training simulators of NPPs (i.e., simulator data). In order to investigate the applicability of these two systems, two papers have been prepared with distinct purposes. The first paper, entitled "SACADA and HuREX: Part 1. The Use of SACADA and HuREX Systems to Collect Human Reliability Data", deals with technical issues pertaining to the collection of HRA data. This second paper explains how the two systems are able to inform HRA practitioners. To this end, the process of estimating HEPs is demonstrated based on feed-and-bleed operations using HRA data from the two systems.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1686-1697
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• 2022

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. Accordingly, HRA community has emphasized the accumulation of HRA data to support HRA practitioners for many decades. To this end, it is critical to resolve practical problems including (but not limited to): (1) how to collect HRA data from available information sources, and (2) how to inform HRA practitioners with the collected HRA data. In this regard, the U.S. Nuclear Regulatory Commission (NRC) and Korea Atomic Energy Research Institute (KAERI) independently initiated two large projects to accumulate HRA data by using full-scale simulators (i.e., simulator data). In terms of resolving the first practical problem, the NRC and KAERI developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. In addition, to inform HRA practitioners, the NRC and KAERI proposed several ideas to extract useful information from simulator data. This paper is the first of two papers to discuss the technical underpinnings of the development of the SACADA and HuREX systems.