• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.772-784
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• 2024

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.303-313
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• 1999

Various interpolation methods have been compared for reconstruction of LMR pin power distributions in hexagonal geometry. Interpolation functions are derived for several combinations of nodal quantities and various sets of basis functions, and tested against fine mesh calculations. The test results indicate that the interpolation functions based on the sixth degree polynomial are quite accurate, yielding maximum interpolation errors in power densities less than 0.5%, and maximum reconstruction errors less than 2% for driver assemblies and less than 4% for blanket assemblies. The main contribution to the total reconstruction error is made tv the nodal solution errors and the comer point flux errors. For the polynomial interpolations, the basis monomial set needs to be selected such that the highest powers of x and y are as close as possible. It is also found that polynomials higher than the seventh degree are not adequate because of the oscillatory behavior.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2422-2430
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• 2020

In order to improve calculational efficiency of the CAPP code in the analysis of the hexagonal reactor core, we have tried to implement a refined AFEN method with transverse gradient basis functions and interface flux moments in the hexagonal geometry. The numerical scheme for the refined AFEN method adopted here is the response matrix method that uses the interface partial currents as nodal unknowns instead of the interface fluxes used in the original AFEN method. Since the response matrix method is single-node based, it has good properties such as good calculational efficiency and parallel computing affinity. Because a refined AFEN method equivalent nonlinear FDM response matrix method tried first could not provide a numerically stable solution, a direct formulation of the refined AFEN response matrix were developed. To show the numerical performance of this response matrix method against the original AFEN method, the numerical error analyses were performed for several benchmark problems including the VVER-440 LWR benchmark problem and the MHTGR-350 HTGR benchmark problem. The results showed a more than three times speedup in computing time for the LWR and HTGR benchmark problems due to good convergence and excellent calculational efficiency of the refined AFEN response matrix method.

• 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).

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.22-22
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• 1999

The HELlOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC "Kurchatov Institute" with water-moderated hexagonally pitched lattices of highly enriched Uranium fuel rods (8Ow/o). We also used the same input by using the MCNP code that was described in the evaluation report, and compared our results with those of the evaluation report. HELlOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves neutronics model with the AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that the HELIOSMASTER code system is fast and accurate enough to be used as nuclear core analysis tool for hexagonal geometry.ometry.

• Journal of the Korean Magnetic Resonance Society
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• v.19 no.1
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• pp.18-22
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• 2015

The structural geometry of $[N(CH_3)_4]_2CdCl_4$ in a hexagonal phase is studied by $^1H$ MAS NMR, $^{13}C$ CP/MAS NMR, and $^{14}N$ NMR. The changes in the chemical shifts for $^{13}C$ and $^{14}N$ in the hexagonal phase are explained by the structural geometry. In addition, the temperature dependencies of the spin-lattice relaxation time in the rotating frame $T_{1{\rho}}$ for $^1H$ MAS NMR and $^{13}C$ CP/MAS NMR are measured.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.142-147
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• 1996

The analytic function expansion nodal (AFEN) method has been successfully applied to two-group neutron diffusion problems. In this paper, the AFEN method is extended to solve general multigroup equations for any type of geometries. Also, a suite of new nodal codes based on the extended AFEN theory is developed for hexagonal-z geometry and applied to several benchmark problems. Numerical results obtained attest to their accuracy and applicability to practical problems.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.15-15
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• 1999

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2003.06a
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• pp.1410-1413
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• 2003

Drawing is a basic plastic deformation method and productive manufacturing process make wire. rod and variety section geometry bar. Study for the rod drawing process of rod was researched long littles. but non-axisymmetric drawing process is weak. So metal flow is very irregular in non-axisymmetric drawing process and difficult to define about material deformation generally. In this paper, to solve material deformation, use finite element method and then define suitable shape for rod to hexagonal drawing dies. And research corner filling rate and surface roughness for the high strength steel hexagonal bar produced defined dies.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.379-384
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• 1996

Thermal-hydraulic performance has been analyzed for an advanced reactor core loaded with hexagonal fuel assemblies. Currently available CHF prediction models and data base for triangular lattice bundles have been thoroughly reviewed, and as a result the KfK-3 CHF correlation with limit CHFR of 1.235 has been determined to be most appropriate. The pressure drop model in COBRA-IV-I code has been modified for the analysis of triangular lattice rod bundles. In view of maximizing the thermal margin, the geometry of a hexagonal fuel assembly, such as rod diameter and rod pitch, has been optimized with a fixed fuel assembly cross sectional area The optimum value of the moderator-to-fuel volume ratio is estimated to lie between 0.65 to 1 with 9.5 mm rod diameter. The thermal margin of these hexagonal fuel assemblies in the AP600 core has been evaluated and compared with that of square lattice fuel assemblies such as VANTAGE-5H and KOFA. The analysis result shows that the performances of hexagonal fuel assemblies are more favorable than the square fuel assemblies in the aspect of steady-state overpower margin.