• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1791-1801
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• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.280-285
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• 2001

This paper presents a modelling of thermal discharge performance for a static ice-on-coil ice-storage tank. Through the present study, discharging characteristics were examined with the existing results of theoretical and numerical heat transfer analyses. Also, an experiment was conducted to obtain a real set of discharge performance. The thermal effectiveness, the ratio of the actual heat transfer rate to the maximum possible heat transfer rate, decreased when the stored energy decreased during discharging period. And the effectiveness increased as the coolant flow rate through the storage increased, of which increasing rate decreased abruptly near the maximum and the minimum stored energy. An empirical correlation was obtained from the experimental and the numerical analysis data.

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.406-416
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• 1997

A computational method is developed for predicting the steady-state temperature field in an LMR core. Detailed core-wide coolant temperature profiles are efficiently calculated using the simplified energy equation mixing model[1] and the subchannel analysis method. The $\theta$-method is employed for discretizing the energy equations in the axial direction. The interassembly coupling is achieved by interassembly gap flow. Cladding and fuel temperatures are calculated with the one-dimensional conduction model and temperature integrals of conductivities. The accuracy of the method is tested by performing several benchmark calculations for too LMR problems. The results indicate that the accuracy is comparable to the other methods based on ENERGY model. It is also shown that the implicit scheme for the axial discretization is more efficient than the explicit scheme.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.31 no.10
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• pp.109-115
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• 1982

Reactor dynamics were studied by reactor noise at TRIGA MARK Il reactor whose rated power is 250KW thermal. The power spectral densities(PSD) of the noise were measured by stochastic method with high resolution digital filters and Fast Fourier Transformers. The transfer function of the reactor at zero power was identical to the theoretical characteristics. When the power was increasec above 1KW, reactor showed its poswer resonances at 3Hz and 10 Hz. It was analyzed that 3Hz peak was generated by heat transfer and coolant flow effects and 10Hz peak by nuclear reaction effects.

• Journal of the Korean Society of Safety
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• v.11 no.2
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• pp.52-59
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• 1996

This paper presents a new approach for assessing the dynamic reliability in a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences which may occur during mid-loop operation. Mid-loop operation is defined as an operation to make RCS water level below the top of the flow area of the hot legs at the junction with the reactor vessel for repairs and maintenance of steam generators and reactor coolant pumps for a specific time. The Idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.35-46
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• 1988

The prediction of clad surface temperatures is important to the design and the safety anlaysis of nuclear reactor. The accurate prediction requires the detailed knowledge of the flow structure and heat transfer, which is complicate due to anisotropic turbulent phenomena. A two-equation model including anisotropic eddy viscosity model is applied to forecast the velocity distribution. And the temperature field is calculated with uniform wall heat flux. The Galerkin's weighted residual finite element method has been used to calculate the turbulent quantities right up to the wall. The numerical results show good agreement with available data and that turbulence anisotropy strongly affects on the mean flow and thus the temperature field. And Nu-P/D correlation is established for sodium coolant in close-packed equilateral triangular bundle in the P/D range of 1.05 to 1.30.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.1
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• pp.55-62
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• 2016

Nuclear fuels are always exposed to hot temperature and high speed coolant flow during the reactor operation. Thus the fuel rod accompanies small amplitude vibration due to the turbulent flow. The random vibration causes friction between the fuel rod and the grid structure which provides the lateral supports. The friction is critical to the fuel rod fretting wear, and it degrades fuel performance when a severe wear is developed. LuGre friction model is introduced in the paper, and the performance was evaluated comparing to the classical Coulomb model. It is shown that the developed friction force considering the Coulomb friction is not enough to stop or delay the motion while the stick-slip can be simulated using LuGre friction model. Numerical solutions of the two dimensional spacer grid cell model with the modern friction are also reviewed, and it is discussed that the new friction model simulates well the nonlinear mechanism.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.