• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2887-2900
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• 2020

Cladding failure in a fuel rod during operation in a BWR is analyzed using a FALCON code-based model. Comparative calculation with a neighbouring, intact rod is presented, as well. A considerable 'hot spot' effect in cladding temperature is predicted with the SLICE-DO model due to a thermal barrier caused by the localized crud deposition. Particularly significant overheating is expected to occur on the affected side of the cladding of the failed rod, in agreement with signs of significant localized crud deposition as revealed by Post Irradiation Examination. Different possibilities (criteria) are checked, and Pellet-Cladding Mechanical Interaction (PCMI) is shown to be one of the plausible potential threats. It is shown that PCMI could lead to discernible concentrated inelastic deformation in the overheated part of the cladding. None of the specific mechanisms considered can be experimentally or analytically identified as an only cause of the rod failure. However, according to the current calculation, a possibility of cladding failure by PCMI cannot be excluded if the crud thickness exceeded 300 ㎛.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.414-422
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• 2021

The aim of this work considers a second order point reactor kinetics model based on the P1 approximation of transport theory, called in this work as P1 point reactor model. The P1 point reactor model implicitly considers the time derivative of the neutron source which has not been thus considered previously. The inverse method to calculate the reactivity in nuclear reactors -chosen because its high accuracy- Matrix Formulation. The numerical results shown that the Matrix Formulation for the reactivity estimation constitutes a method with insignificant calculation errors.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.316-321
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• 1997

The mixing vane effect on the Critical Heat Flux (CHF) is discussed with focus on the vortex now effect. In the subchannel approach, this effect is not quantified by the calculation model, but directly taken into account by the CHF correlation itself through data analysis. The vortex now effect is identified the two Westinghouse correlations, and then the CHF margin issue given rise to by the Vantage-5H design change is evaluated and discussed. It is noted that deficiency about CHF dependency on the vortex flow effect could induce an error in the Departure from Nucleate Boiling Ratio (DNBR) sensitivity Calculation.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.34-45
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• 2001

We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.

• 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.55 no.11
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• pp.4102-4111
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• 2023

In order to better perform thermal hydraulic calculation and analysis of supercritical water reactor, based on the experimental data of supercritical water, the model training and predictive analysis of the heat transfer coefficient of supercritical water were carried out by using the support vector machine (SVM) algorithm. The changes in the prediction accuracy of the supercritical water heat transfer coefficient are analyzed by the changes of the regularization penalty parameter C, the slack variable epsilon and the Gaussian kernel function parameter gamma. The predicted value of the SVM model obtained after parameter optimization and the actual experimental test data are analyzed for data verification. The research results show that: the normalization of the data has a great influence on the prediction results. The slack variable has a relatively small influence on the accuracy change range of the predicted heat transfer coefficient. The change of gamma has the greatest impact on the accuracy of the heat transfer coefficient. Compared with the calculation results of traditional empirical formula methods, the trained algorithm model using SVM has smaller average error and standard deviations. Using the SVM trained algorithm model, the heat transfer coefficient of supercritical water can be effectively predicted and analyzed.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.547-556
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• 2011

Two computational methods for determining equilibrium states for the Mark-IV electrorefiner (ER) have been assessed to improve the current computational electrorefiner model developed at University of Idaho. Both methods were validated against measured data to better understand their effects on the calculation of the equilibrium compositions in the ER. In addition, a sensitivity study was performed on the effect of specific unknown activity coefficients-including sodium in molten cadmium, zirconium in molten cadmium, and sodium chloride in molten LiCl-KCl. Both computational methods produced identical results, which stayed within the 95% confidence interval of the experimental data. Furthermore, sensitivity to unavailable activity coefficients was found to be low (a change in concentration of less than 3 ppm).

• Proceedings of the KIEE Conference
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• 1999.11b
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• pp.450-454
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• 1999

In order to evaluate the integrity of nuclear power plants, J-integral calculation is crucial. For this purpose, finite element method is popularly used to obtain J-integral. However, high cost time consuming preprocess should be performed to design the finite element model of a cracked structure. Also, the J-integral should be verified by alternative method since it may differ depending on the calculation method. The objective of this paper is to develop a three-dimensional elastic-plastic J-integral analysis system which is named as EPAS. The EPAS program consists of an automatic mesh generator for a through-wall crack and a surface crack, a solver based on ABAQUS program, and a J-integral calculation program which provides DI(Domain Integral) and EDI(Equivalent Domain Integral) based J-integral calculation. Using the EPAS program, an optimized finite element model for a cracked structure can be generated and corresponding J-integral can be obtained subsequently.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.379-388
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• 2018

Models for the rate of atomization and deposition of droplets for stratified and annular flow in horizontal pipes are presented. The entrained fraction is the result of a balance between the rate of atomization of the liquid layer that is in contact with air and the rate of deposition of droplets. The rate of deposition is strongly affected by gravity in horizontal pipes. The gravitational settling of droplets is influenced by droplet size: heavier droplets deposit more rapidly. Model calculation and simulation results are compared with experimental data from various diameter pipes. Validation for the suggested models was performed by comparing the Safety and Performance Analysis Code for Nuclear Power Plants calculation results with the droplet experimental data obtained in various diameter horizontal pipes.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.229-236
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• 2004

Radiation hardening is a multiscale phenomenon involving various processes over a wide range of time and length. We present a multiscale model for estimating the amount of radiation hardening in pressure vessel steel in the environment of a light water reactor. The model comprises two main parts: molecular dynamics (MD) simulation and a point defect cluster (PDC) model. The MD simulation was used to investigate the primary damage caused by displacement cascades. The PDC model mathematically formulates interactions between point defects and their clusters, which explains the evolution of microstructures. We then used a dislocation barrier model to calculate the hardening due to the PDCs. The key input for this multiscale model is a neutron spectrum at the inner surface of reactor pressure vessel steel of the Younggwang Nuclear Power Plant No.5. A combined calculation from the MD simulation and the PDC model provides a convenient tool for estimating the amount of radiation hardening.