• 대한안전경영과학회지
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• 제13권2호
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• pp.267-273
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• 2011

Since 'The Act on Private Investment of The Infrastructure' was established in 1994, private investment as well as government's investment in transport infrastructure has been active. However investment in transport infrastructure has more risks than others' due to uncertainty both in traffic volume and in construction cost. In the current appraisal procedure of deciding transportation infrastructure investment, instead of risk management, the sensitivity analysis considering only the changes of benefit, cost and social discount rate which are main factor affecting economic feasibility is carried out. Therefore the uncertainty of various factors affecting demand, cost and benefit are not considered in feasibility study. In this study the problems in current investment appraisal system were reviewed. Using Delphi technique the major factors which have high uncertainty in feasibility study were surveyed and then improvement plan was suggested in the respective of classic 4 step demand forecasting method. The range estimation technique was also mentioned to deal with the uncertainty of the future.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Earthquakes and Structures
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• 제26권4호
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• pp.311-326
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• 2024

Transmission tower structures are particularly susceptible to damage and even collapse under strong seismic ground motions. Conventional seismic analyses of transmission towers are usually performed by considering only ground motion uncertainty while ignoring structural uncertainty; consequently, the performance evaluation and failure prediction may be inaccurate. In this context, the present study numerically investigates the seismic responses and failure mechanism of transmission towers by considering multiple sources of uncertainty. To this end, an existing transmission tower is chosen, and the corresponding three-dimensional finite element model is created in ABAQUS software. Sensitivity analysis is carried out to identify the relative importance of the uncertain parameters in the seismic responses of transmission towers. The numerical results indicate that the impacts of the structural damping ratio, elastic modulus and yield strength on the seismic responses of the transmission tower are relatively large. Subsequently, a set of 20 uncertainty models are established based on random samples of various parameter combinations generated by the Latin hypercube sampling (LHS) method. An uncertainty analysis is performed for these uncertainty models to clarify the impacts of uncertain structural factors on the seismic responses and failure mechanism (ultimate bearing capacity and failure path). The numerical results show that structural uncertainty has a significant influence on the seismic responses and failure mechanism of transmission towers; different possible failure paths exist for the uncertainty models, whereas only one exists for the deterministic model, and the ultimate bearing capacity of transmission towers is more sensitive to the variation in material parameters than that in geometrical parameters. This research is expected to provide an in-depth understanding of the influence of structural uncertainty on the seismic demand assessment of transmission towers.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.313-342
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• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• 한국수자원학회논문집
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• 제45권4호
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• pp.383-391
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• 2012

본 연구는 다기준 의사결정 문제에서 항상 발생하는 가중치와 대안들의 평가치에 대한 불확실성을 최소화하기 위해 민감도 분석을 수행하는 절차를 제시하였다. 제기되는 가중치에 대한 불확실성을 극복하기 위해 일반적으로 순위가 뒤바뀔 수 있는 가장 민감한 평가기준의 결정과 대안의 효과 측정자료의 결정이 있다. 본 연구는 유량확보와 수질개선을 위한 수자원 계획수립을 위해가중합계법을 이용한 문제에 두 경우의 민감도분석을 모두수행하였다. 이 과정에서 결정계수와 민감도 계수를 산정하여 이용하였다. 본 연구에서 제시한 민감도 분석 과정은 향후 수자원 계획 수립에 폭넓게 활용될 수 있다.

• Journal of Radiation Protection and Research
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• 제16권2호
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• pp.67-74
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• 1991

육상 섭취 경로에 따른 내부 피폭선량 계산 모델 KFOOD의 파라메터 불착실성 및 민감도를 몬테칼로법을 사용하여 수치 분석하였다. 쌀을 통한 섭취 경로의 경우 KFOOD 코드에 의한 예측치는 아주 보수적인 값을 나타내었다 모델에서 민감도가 큰 입력변수는 방사능의 침적속도와 식물전이제수였다.

• 대한기계학회논문집A
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• 제25권3호
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• pp.520-528
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• 2001

This paper investigates a resonant microaccelerometer that measures acceleration using a built-in micromechanical resonator, whose resonant frequency is changed by the acceleration-induced axial force. A set of design equations for the resonant microaccelerometer has been developed, including analytic formulae for resonant frequency, sensitivity, nonlinearity and maximum stress. On this basis, the sizes of the accelerometer are designed for the sensitivity of 10$^3$Hz/g in the detection range of 5g, while satisfying the conditions for the maximum nonlinearity of 5%, the minimum shock endurance of 100g and the size constraints placed by microfabrication process. A set of the resonant accelerometers has been fabricated by the combined use of bulk-micromachining and surface-micromachining techniques. From a static test of the cantilever beam resonant accelerometer, a frequency shift of 860Hz has been measured for the proof-mass deflection of 4.3${\pm}$0.5$\mu\textrm{m}$; thereby resulting in the detection sensitivity of 1.10${\times}$10$^3$Hz/g. Uncertainty analysis of the resonant frequency output has been performed to identify important issues involved in the design, fabrication and testing of the resonant accelerometer.

• 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.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.249-265
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• 1987

고준위 방사성폐기물 처분장에 대한 확률론적 위험도 평가를 위해 지금까지 많은 방법들이 제안되어 왔다. 이 계는 많은 불확실성을 갖는 입력 변수들을 갖고 있어서 이 입력변수들에 대해 계산된 위험도 역시 많은 불착실성을 갖는다. 본 논문에서는 이러한 점들을 조직적으로 분석하기 위하여 여러가지 불확실도 및 민감도 분석 방법들이 개발되었고 고준위 폐기물 처분장의 위험도 평가에 적용되었다. 본 논문을 통해 개발된 통계 패키지 SPUSA는 통계적 열여유도 분석, 방사선원 불확실도 분석등 등의 분야에도 사용될 수 있다.