• Steel and Composite Structures
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• v.17 no.6
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• pp.835-850
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• 2014

The fire resistance of composite steel and concrete structures may be determined by using the simplified methods provided in EN 1994-1-2. For the particular situations not covered by the standard, an advanced calculation model might be applied, using special purpose programs for the analysis of structures in fire. The validation of these programs has always been an important issue for software developers, but also for designers and authorities. Clause 4.4.4 from EN 1994-1-2 refers to the validation of the advanced calculation models and states that these models must be validated through relevant test results. The paper presents the calculation of fire resistance of the composite columns in a high-rise building built in Romania, and focusses on the validation of the calculation model (computer program SAFIR), for this particular case. This validation, asked by the Romanian authorities, considers the available experimental results of a fire test, performed on a similar composite steel-concrete column.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2196-2203
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• 2020

A dryout mechanism model for rectangular narrow channels at high pressure conditions is developed by assuming that the Kelvin-Helmholtz instability triggered the occurrence of dryout. This model combines the advantages of theoretical analysis and empirical correlation. The unknown coefficients in the theoretical derivation are supported by the experimental data. Meanwhile, the decisive restriction of the experimental conditions on the applicability of the empirical correlation is avoided. The expression of vapor phase velocity at the time of dryout is derived, and the empirical correlation of liquid film thickness is introduced. Since the CHF value obtained from the liquid film thickness should be the same as the value obtained from the Kelvin-Helmholtz critical stability under the same condition, the convergent CHF value is obtained by iteratively calculating. Comparing with the experimental data under the pressure of 6.89-13.79 MPa, the average error of the model is -15.4% with the 95% confidence interval [-20.5%, -10.4%]. And the pressure has a decisive influence on the prediction accuracy of this model. Compared with the existing dryout code, the calculation speed of this model is faster, and the calculation accuracy is improved. This model, with great portability, could be applied to different objects and working conditions by changing the expression of the vapor phase velocity when the dryout phenomenon is triggered and the calculation formula of the liquid film.

• Journal of Magnetics
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• v.5 no.3
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• pp.85-89
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• 2000

The hysteretic characteristics of giant magnetoimpedance (GMI) profiles have been measured in Co-based amorphous ribbon with various anisotropy angles $\theta_k$, and they have been analyzed by using the Stoner-Wohl-farth model. Two-peaks behavior with a dip near zero field is revealed in the measured GMI profile at 10 MHz irrespective of $\theta_k$. The negligible hysteresis of the field fur the dip is close to the calculation assuming the magnetization jump from a metastable to stable state. However, the hysteretic asymmetry far the angle range of $20^\circ\leq\theta_k < 60^\circ$ is well described by the divergence in the calculation without the magnetization jump. The asymmetry for $\theta_k\geq60^\circ$ may be due to the divergence, but the shapes of measured profiles are quite different from the calculations with single peak near zero field, indicating that Stoner-Wohlfarth model can be well used to describe GMI characteristics for the anisotropy angle range of $20^\circ\leq\theta_k < 60^\circ$at the frequency of 10 MHz in Co-based amorphous ribbons.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.301-306
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• 1996

To predict the average liquid level under the condition of the countercurrent stratified two-phase flow in a pipe, an analytical model has been suggested. This is made by introducing the interfacial level gradient into the liquid-phase and the gas-phase momentum equations. The analytical method for the gas-phase pressure drop calculation with f$_i$ $\neq$ f$_G$ has also been described using the liquid level prediction model developed in the present study.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1769-1780
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• 2022

A new method of dose calculation algorithm, called GPU-accelerated Monte Carlo and collapsed cone Convolution (GMCC) was developed to improve the calculation speed of BNCT treatment planning system. The GPU-accelerated Monte Carlo routine in GMCC is used to simulate the neutron transport over whole energy range and the Collapsed Cone Convolution method is to calculate the gamma dose. Other dose components due to alpha particles and protons, are calculated using the calculated neutron flux and reaction data. The mathematical principle and the algorithm architecture are introduced. The accuracy and performance of the GMCC were verified by comparing with the FLUKA results. A water phantom and a head CT voxel model were simulated. The neutron flux and the absorbed dose obtained by the GMCC were consistent well with the FLUKA results. In the case of head CT voxel model, the mean absolute percentage error for the neutron flux and the absorbed dose were 3.98% and 3.91%, respectively. The calculation speed of the absorbed dose by the GMCC was 56 times faster than the FLUKA code. It was verified that the GMCC could be a good candidate tool instead of the Monte Carlo method in the BNCT dose calculations.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.9-16
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• 1999

A 330 MWt-sized multi-purpose integral-type reactor, SMART is under development in Korea for the use of nuclear energy other than electricity generation. In this study, various radioactive source terms are estimated for SMART. SMART is different from conventional reactor concepts in operation and design. Therefore Specific Calculation method namely recurrence model is used. This model is based on the change rate in the RC radioactivity materials and operational characteristics of SMART Calculation results show tremendously increase of the levels of RC activity because no cleanup of RC and long term operation.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.790-802
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• 2022

Large-scale reactor simulation often requires the use of Monte Carlo calculation techniques to estimate important reactor parameters. One drawback of these Monte Carlo calculation techniques is they inevitably result in some uncertainty in calculated quantities. The present study includes parametric uncertainty quantification (UQ) and sensitivity analysis (SA) on the Advanced Test Reactor Critical (ATRC) facility housed at Idaho National Laboratory (INL) and addresses some complications due to Monte Carlo uncertainty when performing these analyses. This approach for UQ/SA includes consideration of Monte Carlo code uncertainty in computed sensitivities, consideration of uncertainty from directly measured parameters and a comparison of results obtained from brute-force Monte Carlo UQ versus UQ obtained from a surrogate model. These methodologies are applied to the uncertainty and sensitivity of k_eff for two sets of uncertain parameters involving fuel plate geometry and fuel plate composition. Results indicate that the less computationally-expensive method for uncertainty quantification involving a linear surrogate model provides accurate estimations for k_eff uncertainty and the Monte Carlo uncertainty in calculated k_eff values can have a large effect on computed linear model parameters for parameters with low influence on k_eff.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.636-643
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• 1995

The accurate determination of the fuel-cladding gap conductance as functions of rod burnup and power level may be a key to the design and safety analysis of a reactor. The incorporation of a sophisticated gap conductance model into nuclear design code for computing thermal hydraulic feedback effect has not been implemented mainly because of computational inefficiency due to complicated behavior of gap conductance. To avoid the time-consuming iteration scheme, simplification of the gap conductance model is done for the current design model. The simplified model considers only the heat conductance contribution to the gap conductance. The simplification is made possible by direct consideration of the gas conductivity depending on the composition of constituent gases in the gap and the fuel-cladding gap size from computer simulation of representative power histories. The simplified gap conductance model is applied to the various fuel power histories and the predicted gap conductances are found to agree well with the results of the design model.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.