• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• Journal of Advanced Marine Engineering and Technology
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• v.34 no.4
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• pp.491-499
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• 2010

Heat transfer performance of tube bundles have long been investigated since they were widely used. Most of previous experimental and numerical works for tube bundles were performed with tube diameter in the range of 25~51mm and Reynolds number of $8.000{\leq}Re{\leq}30.000$. Recently, tube bundles with small diameter tube collects interests since the mini-channel tube provides higher compactness. The present work aims to investigate the applicability of previous correlations available in the open literature to the tube bundles with small diameter of 1.5mm and $3.000{\leq}Re{\leq}7.000$. A commercial CFD package was used to analyze the thermal-hydraulic performance of them. The results show that the Zukauskas correlation developed for larger diameter tube and higher Reynolds number are still in good agreement with them within the discrepancy of 4.7%. The analyses also show that the Nuselt number increases with a decrease in the longitudinal pitch.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.15 no.9 s.102
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• pp.1009-1015
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the $PIAT^{(R)}$ (program for integrity assessment of steam generator tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support Plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-nitration bars such as vortical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.135-147
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• 2016

The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.

• Transactions of the Korean hydrogen and new energy society
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• v.19 no.5
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• pp.367-376
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• 2008

To obtain the data of the pressure loss and differential pressure at the inside of the stack that was composed of 126 cells with 7,500 cm2 electrode area, 75kW molten carbonate fuel cell system has been operated. Computational fluid dynamics was applied to estimate reactions and thermal fluid behavior inside of the stack that was adopted with internal manifold type separator. The pressure loss coefficient K showed 72.29 to 84.01 in anode and 6.34 to 8.75 in cathode at low part of cells at the inside of 75 kW MCFC stack respectively. Meanwhile, the pressure loss coefficient of the higher part of cells at the interior of the stack showed 15.36 and 56.44 in anode and cathode respectively. These results mean that there is no big total pressure difference between anode and cathode at the inner part of 75 kW MCFC stack. This result will be reflected in 250kW MCFC system design.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.476-485
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• 1999

In order to evaluate the gravity-injection capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Yong Gwang Units 3&4 were reviewed. The six cases of possible gravity-injection paths from the refueling water tank (RWT) were identified and the thermal-hydraulic analyses were performed using the RELAP5/MOD3.2 code. The core cooling capability was significantly dependent on the gravity-injection path, the RCS opening, and the injection rate. In the cases with the pressurizer manway opening higher than the RWT water level, the coolant was held up in the pressurizer and the system pressure continued increasing after gravity-injection. The gravity injection eventually stopped due to the high system pressure and the core was uncovered. In the cases with the injection path and opening on the same leg side, the core cooling was dependent on whether the water injected from the RWT passed the core region or not. However, in the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. In addition, from the sensitivity study on the gravity-injection flow rate, it was found that about 54 kg/s of injection rate was required to maintain the core cooling and the core cooling could be provided for about 10.6 hours after event with that injection rate from the RWT. Those analysis results would provide useful information to operators coping with the event.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.446-454
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• 2021

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.227-234
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• 2019

A user-friendly modeling interface is developed for a process-based total system performance assessment framework (APro) specialized for a generic geological disposal system for high-level radioactive waste. The APro modeling interface is constructed using MATLAB, and the operator splitting scheme is used to combine COMSOL for simulation of multiphysics and PHREEQC for the calculation of geochemical reactions. As APro limits the modeling domain to the generic disposal system, the degree of freedom of the model is low. In contrast, the user-friendliness of the model is improved. Thermal, hydraulic, mechanical and chemical processes considered in the disposal system are modularized, and users can select one of multiple modules: "Default process" and multi "Alternative process". APro mainly consists of an input data part and calculation execution part. The input data are prepared in a single EXCEL file with a given format, and the calculation part is coded using MATLAB. The final results of the calculation are created as an independent COMSOL file for further analysis.