• 한국수자원학회:학술대회논문집
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• 한국수자원학회 2010년도 학술발표회
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• pp.1390-1394
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• 2010

Dempster-Shafer 이론은 미지의 매개변수 추정시 베이지안 기법의 제약을 완화시키기 위한 베이지안 접근법의 일반화로 해석될 수 있으며, 상호배타적인 싱글톤에만 확률이 할당되는 것이 아니라 가능한 결과의 부분집합들이 기본확률할당을 위한 대상으로 고려된다. 베이지안 접근은 우연적 불확실성 및 지식의 불확실성을 효율적으로 구분할 수 없으며, 특정도가 낮고 애매한 증거들을 다룰 수 없는 반면, Dempster-Shafer 증거추론은 이러한 문제들을 효율적으로 평가할 수 있다. 따라서 본 논문에서는 홍수위험평가 및 수자원 계획 수립시 가장 기본이 되는 강우빈도해석에서 확률분포의 매개변수에 대한 불확실성 고려한 확률강우량의 산정 및 불확실성의 영향을 평가하기 위하여 Dempster-Shafer 이론을 이용하여 불확실성을 고려한 강우빈도해석모델 구축 및 적용을 통해 홍수위험평가 및 수자원 계획 등에 있어서 불확실성 표현 및 처리기법을 제시하였다.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1219-1225
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• 2017

In this work, a scalable algorithm for model calibration in nuclear engineering applications is presented and tested. The algorithm relies on the construction of surrogate models to replace the original model within the region of interest. These surrogate models can be constructed efficiently via reduced order modeling and subspace analysis. Once constructed, these surrogate models can be used to perform computationally expensive mathematical analyses. This work proposes a surrogate based model calibration algorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulics parameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used to simulate a three-dimensional core depletion problem. The proposed algorithm is then used to construct a reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronics and thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilation and calibration are highlighted in an uncertainty quantification study and requantification after the calibration process. Results showed that the proposed algorithm could help to reduce the uncertainty in key reactor attributes based on experimental and operational data.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1687-1707
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• 2023

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann's model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon's 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3787-3800
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• 2023

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.389-400
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• 1994

FLECHT SEASET 실험 데이터를 사용하여 대형 냉각재 상실 사고시 재관수 첨두 피복재 온도에 대한 RELAP5 /MOD3/KAERI의 예측능력을 평가하였으며, 관련 불확실성을 통계적으로 정량화 하였다. 중력구동 재관수 실험및 광범위한 재관수율, 시스템 압력, 초기 피복재 온도, 연료봉 출력을 포괄하는 강제구동 재관수 실험들로 구성된 18개의 실험이 평가에 사용되었다. 평가 결과 재관수 첨두 피복재 온도에 대해 평균 7.56 K 낮게 예측하였으며 이를 포함한 관련 불확실성의 상한은 95% 신뢰도 수준에서 약 99 K로 정량화 되었다.

• Nuclear Engineering and Technology
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• 제28권6호
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• pp.507-521
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• 1996

Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current Probabilistic safety assessment. The main objective of the present study is to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a Hypothetical severe accident sequence of a station blackout at Younggwang nuclear power plant using MAAP3. OB code as a benchmark problem. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficient (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results that can be obtained by the combined procedure proposed in the present work.

• Wind and Structures
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• 제21권5호
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• pp.461-487
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• 2015

In recent years, the Graphics Processing Unit (GPU) has become a competitive computing technology in comparison with the standard Central Processing Unit (CPU) technology due to reduced unit cost, energy and computing time. This paper describes the derivation and implementation of GPU-based algorithms for the analysis of wind loading uncertainty on high-rise systems, in line with the research field of probability-based wind engineering. The study begins by presenting an application of the GPU technology to basic linear algebra problems to demonstrate advantages and limitations. Subsequently, Monte-Carlo integration and synthetic generation of wind turbulence are examined. Finally, the GPU architecture is used for the dynamic analysis of three high-rise structural systems under uncertain wind loads. In the first example the fragility analysis of a single degree-of-freedom structure is illustrated. Since fragility analysis employs sampling-based Monte Carlo simulation, it is feasible to distribute the evaluation of different random parameters among different GPU threads and to compute the results in parallel. In the second case the fragility analysis is carried out on a continuum structure, i.e., a tall building, in which double integration is required to evaluate the generalized turbulent wind load and the dynamic response in the frequency domain. The third example examines the computation of the generalized coupled wind load and response on a tall building in both along-wind and cross-wind directions. It is concluded that the GPU can perform computational tasks on average 10 times faster than the CPU.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• 한국식품과학회지
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• 제53권4호
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• pp.391-398
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• 2021

As global interest in herbal medicines has increased, the adulteration of herbal medicines has become a critical safety issue. Adulteration with dyes to improve the appearance of low-quality products is of particular concern. This study aimed to develop a high-performance liquid chromatography (HPLC) method to detect dyes added as adulterants to Phellodendron. Samples were analyzed on a C18 column using 50 mM ammonium acetate and acetonitrile as the mobile phase. All calibration curves showed good linearity (r² ≥0.9999) over the five-point concentration range (1-50 mg/kg). Limit of detection ranged from 0.04-0.35 mg/kg, and limit of quantification ranged from 0.11-1.07 mg/kg. The repeatability and reproducibility for these measurements were 94.2-103.3% and 96.6-103.8% for accuracy and 0.14-2.28 RSD (%) and 0.80-2.37 RSD (%) for precision. Moreover, the measurement uncertainty of the low, medium, and high concentrations for 10 dyes was considered. Thus, this HPLC method is suitable for detecting color adulteration of Phellodendron.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2151-2161
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• 2021

Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.