• Structural Engineering and Mechanics
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• v.50 no.6
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• pp.755-772
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• 2014

A simplistic approach towards evaluation of complete load deflection response of Reinforced Concrete (RC) flexural members under post fire (residual) scenario is presented in this paper. The cross-section of the RC flexural member is divided into a number of sectors. Thermal analysis is performed to determine the temperature distribution across the section, for given fire duration. Temperature-dependent stress-strain curves for concrete and steel are then utilized to perform a moment-curvature analysis. The moment-curvature relationships are obtained for beams exposed to different fire durations. These are then utilized to obtain the load-deflection plots following pushover analysis. Moreover one of the important issues of modeling the initial stiffness giving due consideration to stiffness degradation due to material degradation and thermal cracking has also been addressed in a rational manner. The approach is straightforward and can be easily programmed in spreadsheets. The presented approach has been validated against the experiments, available in literature, on RC beam subjected to different fire durations viz. 1hr, 1.5hrs and 2hrs. Complete load-deflection curves have been obtained and compared with experimentally reported counterparts. The results also show a good match with the results obtained using more complicated approaches such as those involving Finite element (FE) modeling and conducting a transient thermal stress analysis. Further evaluation of the beams during fire (at elevated temperatures) was performed and a comparison of the mechanical behavior of RC beams under post fire and during fire scenarios is made. Detailed formulations, assumptions and step by step approach are reported in the paper. Due to the simplicity and ease of implementation, this approach can be used for evaluation of global performance of fire affected structures.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.467-472
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• 1996

This study discusses special aspects of severe accident related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor.$^{1, 2)}$ The analytical comparison of three elements core to former two elements case is conducted including evaluation of suitable nuclear cross-section sets to account for the effects of system configulation, fuel and moderator mixture temperature, material dispersion and the other thermal-hydraulics. Three elements core ANS reactor is the alternative core design which was proposed as a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium (former uranium fuel is the baseline design value of 93%) A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies still on geometry, material constituents, and thermal-hydraulic conditions are verified. Therefore, the concepts of mitigative design features are qualified.d.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.111-120
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• 1980

Nuclear physics parameters of the Wolsung CANDU-PHW reactor are computed by use of the PHWCELL computer code that is an improved version of LATREP. The PHWCELL code mainly computes cell parameters of heavy water moderated reactors, and modeling scheme of heavy water reactor cell calculations has been developed with the PHWCELL computer code. The reactor operating conditions considered in the study are cold zero power (CZP) and hot full power (HFP) with equilibrium poison. The cell parameters are also computed as a function of fuel burnup and the numerical results are compared with the results in PSR of the Wolsung unit and in the previous study.

• 제어로봇시스템학회:학술대회논문집
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• 1994.10a
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• pp.207-212
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• 1994

A mathematical model is developed for a batch reactor in which the free radical bulk copolymerization of styrene and acrylonitrile takes place. In this model, we introduce the free volume theory to quantify the diffusion controlled termination and propagation reactions, and develop a model for the chain length dependent termination reaction in the context of the pseudo kinetic rate constant method(PKRCM). The simulation results from this model are found to be in good agreement with experimental data under different copolymerization conditions. The present model can predict both the copolymer composition and the number and weight average molecular weights. These kinetic approaches provide greater insight into the performance of the batch reactor used for the free radical bulk copolymerization of styrene and acrylonitirle.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.451-451
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• 2000

This study aims for the optimization of process conditions in a fixed-bed catalytic reactor system with a circulating molten salt bath, in which partial oxidation of propylene to acrolein takes place. Two-dimensional pseudo-homogeneous model is adopted with estimation of suitable parameters and its validity is corroborated by comparing simulation result with experimental data. The temperature of the molten salt and the feed composition are found to exercise significant influence on the yield of acrolein and the magnitude of hot spot. The temperature of the molten salt is usually kept constant. This study, however, suggests that the temperature of the molten salt must be axially adjusted so that the abrupt peak of hot spot should not appear near the reactor entrance. The yield of acrolein is maximized and the position and the magnitude of hot spot are optimized by the method of the iterative dynamic programming (IDP).

• Journal of Korean Society of Water and Wastewater
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• v.13 no.3
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• pp.42-55
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• 1999

A mathematical model for biological nutrient removal in a sequencing batch reactor process, which is based on the IAWQ Activated Sludge Model No. 2 with a few modifications, has been developed. Twenty water quality components and twenty three kinetic equations are incorporated in the model. The model is structured in the matrix form based on the law of mass conservation using stoichiometry and kinetic equations. Stoichiometric coefficients and kinetic parameters included in the model equations are chosen from the literature. A multistep predictor-corrector algorithm of variable step-size is adopted for solving the vector nonlinear ordinary differential equations. The simulation for experimental results is conducted to evaluate the validity of the model and to calibrate coefficients and parameters. The simulation using the model well represents the experimental results from laboratory. The mathematical model developed in this study may be utilized for the design and operation of a sequencing batch reactor process under the steady and unsteady-state at various environmental conditions.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.269-274
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• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1277-1288
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• 2018

In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.