• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.91-96
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• 1995

In this paper, research efforts to develop computer code system for analysis of the transmutation reactors at KAIST are described Especially the computer code HANCELL for assembly calculation of fast reactors is mainly described. Features and function of the code are identified md current status of the code development is provided

• Journal of Energy Engineering
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• v.24 no.1
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• pp.149-163
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• 2015

In this study, we conducted study on the confirmation of thermal-hydraulic safety for Mast assembly with Computational Fluid Dynamics(CFD) analysis. Before performing the natural convection analysis for the Mast assembly by using CFD code, we validated the CFD code against two benchmark natural convection data for the evaluation of turbulence models and confirmation of its applicability to the natural convection flow. From the first benchmark test which was performed by Betts et al. in the simple rectangular channel, we selected standard k-omega turbulence model for natural convection. And then, calculation performance of CFD code was also investigated in the sub-channel of rod bundle by comparing with PNL(Pacific Northwest Laboratory) experimental data and prediction results by MATRA and Fluent 12.0 which were performed by Kwon et al.. Finally, we performed main natural convection analysis for fuel assembly inside the Mast assembly by using validated turbulence model. From the calculation, we observed stable natural circulation flow between the mast assembly and pool side and evaluated the thermal-hydraulic safety by calculating the departure from nucleate boiling ratio.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.11
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• pp.1239-1243
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• 2009

To evaluate the structural integrity of the nuclear seismic category penetration cooling water pump under the seismic service conditions the seismic analysis was performed in accordance with IEEE-STD-344 code. The finite element computer program, ANSYS, Version 10.0, is used to perform both a mode frequency analysis and an equivalent static seismic analysis of the pump assembly. The mode frequency analysis results show the fundamental natural frequency is greater than 33 Hz and does not exist in seismic range, thus justifying the use of the static analysis. The stresses resulted from various loadings and their combinations are within the allowable limits specified in the above mentioned IEEE code. The results of the seismic evaluation fully satisfied the structural acceptance criteria of the IEEE code. Accordingly the structural integrity on the pump assembly was proved.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.5
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• pp.488-496
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• 2003

To establish nondestructive test method for rocket motor assembly with rubber and aerospace composite materials, practicable quality control acoustic emission test method is presented. Structural analysis for motor assembly is performed by ABAQUS code and analysis output result is confirmed by strain gage and AE data. Various specimens were tested and analyzed using strain gage and acoustic emission data. The hit rate of acoustic emission was closely related with case/rubber debonding. This report also describes practicable acoustic emission nondestructive method for evaluating motor case assembly quality assurance in the industrial field.

• Proceedings of the Korean Reliability Society Conference
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• 2000.11a
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• pp.87-95
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• 2000

The testing product is the HSU(Hydrostatic Steering Unit) installed in the armored vehicle which has the compact structure of assembly with hydraulic pump, hydraulic motor, and several hydraulic valves. There are no assesment testing code for HSU within a country because the testing code for HUS is not involved in the technical concert from USA. Therefore the testing code is developed by the dynamic analysis of the armored vehicle. By the developed testing code, the test equipment is designed, manufactured, and used to apply the assesment test.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.128-128
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• 2001

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1280-1286
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• 2023

MgF₂ as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF₂ were calculated, based on the phonon expansion model. The structural properties of MgF₂ were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF₂. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF₂ were reduced by 27.7% and 8.2%, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.28-33
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• 1996

A transport theory code CRX-Hex based on characteristic methods with a general geometric tracking routine is developed for the heterogeneous hexagonal geometry. With the general geometric tracking routine, the formulation of the characteristic method is not changed. To test the code, it was applied to two benchmark problems which consist of complex meshes and compared with other codes (HELIOS, TWOHEX).

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.