• Proceedings of the KSME Conference
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• 2002.08a
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• pp.43-46
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• 2002

The HANARO, multi-purpose research reactor, 30 MWth open-tank-in-pool type, is under 24 MWth of power operation since it reached to the initial critical in February, 1995. Many useful experiments should be safely performed to activate the utilization of the HANARO, but there is a radioactive risk of using the HANARO. To reduce the risk, a test facility, which is not reacted by nuclear fuel, is being developed to simulate similar flow characteristics with the HANARO. This paper describes the computational flow analysis to determine each shape of simulating fuels for simulating the flow similarities of 36 elements hexagonal fuels assembly and 18 elements circulating fuels assembly loaded in HANARO. The shares of orifices were determined by the trial and error method and the structural integrities of them were verified by the finite element method assuming that the flow rate and pressure differences of reactor core are constant. The analysis results will be verified with the results of the flow test to be performed after the installation of this test facility.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.840-845
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• 2004

The remote operation of the Advanced Spent Fuel Conditioning Process (ACP) is analyzed by using the 3D graphic simulation tools. Since the spent nuclear fuel, which is a high radioactive material, is processed in the ACP, the ACP equipment is operated in intense radiation fields as well as in a high temperature. Thus, the equipment is operated in a remote manner and should be designed with consideration for the remote handling and maintenance. Also suitable remote handling technology needs to be developed along with the design of the process concepts. For this we developed a graphic simulator, which provides the capability of verifying the remote operability of the ACP without the fabrication of the process equipment. In other words, by applying virtual reality to the remote maintenance operation, a remote operation task can be simulated in the graphic simulator, not in the real environment. The graphic simulator will substantially reduce the cost of the development of the remote handling and maintenance procedure as well as the process equipment, while at the same time developing a remote maintenance concept that is more reliable, easier to implement, and easier to understand.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.12
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• pp.1653-1661
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• 2012

In this study, the pressure drop changes and structural characteristics of a SMART rod bundle under the effect of a coolant were investigated. The turbulence model of the BSL Reynolds stress model was used to model the coolant flow, and a fluid solid interaction simulation was conducted. First, fuel rod vibration analysis was performed to confirm the natural frequency of the fuel rod, which was supported by spacer grid assemblies, and this was compared with experimental results. From the experimental results, the natural frequency was found to be 48 Hz, and the error compared with the simulation results was 2%. The pressure drop at the rod bundle was calculated and compared with the experimental data; it showed an error of 8%, demonstrating the simulation accuracy. In the flow analysis, the flow velocity and secondary flow at different domains were calculated, and vortex generation was also observed. Finally, through the fluid solid interaction analysis, the fuel rod displacements caused by flow-induced vibrations were calculated. Then, calculated displacement PSD at maximum displacement happed point.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2937-2952
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• 2021

A typical three-dimensional finite element model for a fuel assembly is established, which is composed of 16 monolithic U-10Mo fuel plates and Al alloy frame. The distribution and evolution results of temperature, displacement and stresses/strains in all the parts are numerically obtained and analyzed with a self-developed code of FUELTM. The simulation results indicate that (1) the out-of-plane displacements of Al alloy side plates are mainly attributed to the bending deformations; (2) enhanced out-of-plane displacements appear in fuel plates adjacent to the outside Al plates, which results from the occurred bending deformations due to the applied constraints of outside Al plates; (3) an intense interaction of fuel foil with the cladding occurs near the foil edge, which appears more heavily in the fuel plates adjacent to the outside Al plates. The maximum first principal stresses in the fuel foil are similar for all the fuel plates and appear near the fuel foil edge; while, the through-thickness creep strains of fuel foil in the fuel plate near the central region of fuel assembly are larger, and the induced creep damage might weaken the fuel skeleton strength and raise the fuel failure risk.

• 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.28 no.3
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• pp.331-338
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• 1996

In the thermal-hydraulic design, the effect of fuel rod bow is quantified tv the rod bow DNBR penalty which is a key design parameter to assure the coolability of fuel assembly in the pressurized water reactor. In this work, a computer program for the evaluation of the rod bow DNBR penalty based on Westinghouse methodology is developed and its application procedure is proposed. The computer simulation is based on the Monte-Carlo method. The qualification of developed computer program is performed by a comparison of calculational result with that given by Westinghouse's document. A new application procedure is built using batch mean and batch standard deviation. The normality of sample population generated by the batch calculation is confirmed by means of a chi-square test for goodness of fit. On the view point of statistics it is effected that the more reliable design value may be produced by the new application procedure.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3264-3274
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• 2021

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.