• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4534-4550
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• 2022

Integrated severe accident codes should be capable of simulating not only specific physical phenomena but also entire plant behaviors, and in a sufficiently fast time. However, significant uncertainty may exist owing to the numerous parametric models and interactions among the various phenomena. The primary objectives of this study are to present best-practice uncertainty and sensitivity analysis results regarding the evolutions of severe accidents (SAs) and fission product source terms and to determine the effects of mitigation measures on them, as expected during a short-term station blackout (STSBO) of a reference pressurized water reactor (optimized power reactor (OPR)1000). Three reference scenarios related to the STSBO accident are considered: one base and two mitigation scenarios, and the impacts of dedicated severe accident mitigation (SAM) actions on the results of interest are analyzed (such as flammable gas generation). The uncertainties are quantified based on a random set of Monte Carlo samples per case scenario. The relative importance values of the uncertain input parameters to the results of interest are quantitatively evaluated through a relevant sensitivity/importance analysis.

• The KSFM Journal of Fluid Machinery
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• v.14 no.1
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• pp.42-47
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• 2011

Measurement uncertainty analysis of fuel flow using turbine flowmeter was performed for the case of altitude engine test. SAE ARP4990 was used as the fuel flow calculation procedure, as well as the mathematical model for the measurement uncertainty assessment. The assessment was performed using Sensitivity Coefficient Method. 11 parameters involved in the calculation of the flow rate were considered. For the given equipment setup, the measurement uncertainty of fuel flow was assessed in the range of 1.19~1.86 % for high flow rate case, and 1.47~3.31 % for low flow rate case. Fluctuation in frequency signal from the flowmeter had the largest influence on the fuel flow measurement uncertainty for most cases. Fuel temperature measurement had the largest for the case of low temperature and low flow rate. Calibration of K-factor and the interpolation of the calibration data also had large influence, especially for the case of very low temperature. Reference temperature, at which the reference viscosity of the sample fuel was measured, had relatively small contribution, but it became larger when the operating fuel temperature was far from reference temperature. Measurement of reference density had small contribution on the flow rate uncertainty. Fuel pressure and atmospheric pressure measurement had virtually no contribution on the flow rate uncertainty.

• Journal of the Society of Naval Architects of Korea
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• v.49 no.1
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• pp.60-67
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• 2012

This study considers the numerical analysis on parametric roll for container ships. As a method of numerical simulation, an impulse-response-function approach is applied in time domain. A systematic study is carried out for the parametric roll of two container ships, particularly observing the sensitivity of computational results to some parameters which can affect the analysis of parametric roll. The parameters to be considered are metacentric height (GM), simulation time window, and the discretization of wave spectrum. Based on the result of parametric roll simulation, numerical sensitivity and uncertainty in computational analysis are discussed.

• Journal of Korean Society for Quality Management
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• v.45 no.3
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• pp.533-547
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• 2017

Purpose: This study proposes the main sources of uncertainty and uncertainty analysis of a measurement system of insensitive munitions tests. Methods: We established the mathematical model for calculating measurement uncertainty of insensitive munitions tests, conducted experiments for calculating uncertainties of dynamic sensitivity and overshoot value, and estimated the distributions of uncertainty factors. Results: The measurement uncertainty calculation methods are presented, which include experimental data processing methods for calculating uncertainties of dynamic sensitivity and overshoot value. Conclusion: The measurement of explosion pressure in insensitive munitions tests is an important issue to the reporting test results and classifying reaction types. The more efforts to ensure the reliability of the insensitive munitions tests results are required.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

• Earthquakes and Structures
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• v.17 no.2
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• pp.131-141
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• 2019

Structural control systems have uncertainties in their structural parameters and control devices which by using reliability analysis, uncertainty can be modeled. In this paper, reliability of controlled structures equipped with semi-active Magneto-Rheological (MR) dampers is investigated. For this purpose, at first, the effect of the structural parameters and damper parameters on the reliability of the seismic responses are evaluated. Then, the reliability of MR damper force is considered for expected levels of performance. For sensitivity analysis of the parameters exist in Bouc- Wen model for predicting the damper force, the importance vector is utilized. The improved first-order reliability method (FORM), is used to reliability analysis. As a case study, an 11-story shear building equipped with 3 MR dampers is selected and numerically obtained experimental data of a 1000 kN MR damper is assumed to study the reliability of the MR damper performance for expected levels. The results show that the standard deviation of random variables affects structural reliability as an uncertainty factor. Thus, the effect of uncertainty existed in the structural model parameters on the reliability of the structure is more than the uncertainty in the damper parameters. Also, the reliability analysis of the MR damper performance show that to achieve the highest levels of nominal capacity of the damper, the probability of failure is greatly increased. Furthermore, by using sensitivity analysis, the Bouc-Wen model parameters which have great importance in predicting damper force can be identified.

• Journal of the Korean Institute of Gas
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• v.5 no.4 s.16
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• pp.70-78
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• 2001

In order to find out which equipment failures were mostly contributed to the rupture of a continuous stirred tank reactor, the sensitivity analysis was carried out. The uncertainty of likelihood of the rupture of reactor was studied by the uncertainty analysis. And the cost effectiveness analysis resulted in the recommendation of the exchange with a better reliable unit if you want to maintain the process efficiently from the view point of cost. The uncertainty analysis showed that the likelihood of catastrophic rupture of the reactor was distributed from $8.09{\times}10^{-04}$ to $5.50{\times}10^{-02}/year$. As a result of cost-effectiveness analysis, it was proposed to exchange the voting logic unit for a better safer system.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4272-4279
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• 2022

In this study, a cross section stochastic sampling (S.S.) capability is implemented into both the McCARD continuous energy Monte Carlo code and MIG multiple-correlated data sampling code. The ENDF/B-VII.1 covariance data based 30 group cross section sets and the SCALE6 covariance data based 44 group cross section sets are sampled by the MIG code. Through various uncertainty quantification (UQ) benchmark calculations, the McCARD/MIG results are verified to be consistent with the McCARD stand-alone sensitivity/uncertainty (S/U) results and the XSUSA S.S. results. UQ analyses for Three Mile Island Unit 1, Peach Bottom Unit 2, and Kozloduy-6 fuel pin problems are conducted to provide the uncertainties of k_eff and microscopic and macroscopic cross sections by the McCARD/MIG code system. Moreover, the SNU S/U formulations for uncertainty propagation in a MC depletion analysis are validated through a comparison with the McCARD/MIG S.S. results for the UAM Exercise I-1b burnup benchmark. It is therefore concluded that the SNU formulation based on the S/U method has the capability to accurately estimate the uncertainty propagation in a MC depletion analysis.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.