• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3059-3072
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• 2022

A Jacobian-Free Newton Krylov Two-Nodal Coarse Mesh Finite Difference algorithm based on Nodal Expansion Method (NEM_TNCMFD_JFNK) is successfully developed and proposed to solve the three-dimensional (3D) and multi-group reactor physics models. In the NEM_TNCMFD_JFNK method, the efficient JFNK method with the Modified Incomplete LU (MILU) preconditioner is integrated and applied into the discrete systems of the NEM-based two-node CMFD method by constructing the residual functions of only the nodal average fluxes and the eigenvalue. All the nonlinear corrective nodal coupling coefficients are updated on the basis of two-nodal NEM formulation including the discontinuity factor in every few newton steps. All the expansion coefficients and interface currents of the two-node NEM need not be chosen as the solution variables to evaluate the residual functions of the NEM_TNCMFD_JFNK method, therefore, the NEM_TNCMFD_JFNK method can greatly reduce the number of solution variables and the computational cost compared with the JFNK based on the conventional NEM. Finally the NEM_TNCMFD_JFNK code is developed and then analyzed by simulating the representative PWR MOX/UO₂ core benchmark, the popular NEACRP 3D core benchmark and the complicated full-core pin-by-pin homogenous core model. Numerical solutions show that the proposed NEM_TNCMFD_JFNK method with the MILU preconditioner has the good numerical accuracy and can obtain higher computational efficiency than the NEM-based two-node CMFD algorithm with the power method in the outer iteration and the Krylov method using the MILU preconditioner in the inner iteration, which indicates the NEM_TNCMFD_JFNK method can serve as a potential and efficient numerical tool for reactor neutron diffusion analysis module in the JFNK-based multiphysics coupling application.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Korean Journal of Materials Research
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• v.20 no.11
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• pp.570-574
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• 2010

A diamond-like carbon (DLC) film deposited on a WC disk was investigated to improve disk wear resistance for injection molding of zirconia optical ferrule. The deposition of DLC films was performed using the filtered vacuum arc ion plating (FV-AIP) system with a graphite target. The coating processing was controlled with different deposition times and the other conditions for coating, such as input power, working pressure, substrate temperature, gas flow, and bias voltage, were fixed. The coating layers of DLC were characterized using FE-SEM, AFM, and Raman spectrometry; the mechanical properties were investigated with a scratch tester and a nano-indenter. The friction coefficient of the DLC coated on the WC was obtained using a pin-on-disk, according to the ASTM G163-99. The thickness of DLC films coated for 20 min. and 60 min. was about 750 nm and 300 nm, respectively. The surface roughness of DLC films coated for 60 min. was 5.9 nm. The Raman spectrum revealed that the G peak of DLC film was composed of $sp^3$ amorphous carbon bonds. The critical load (Lc) of DLC film obtained with the scratch tester was 14.6 N. The hardness and elastic modulus of DLC measured with the nano-indenter were 36.9 GPa and 585.5 GPa, respectively. The friction coefficient of DLC coated on WC decreased from 0.2 to 0.01. The wear property of DLC coated on WC was enhanced by a factor of 20.

• Nuclear Engineering and Technology
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• v.40 no.2
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• pp.117-126
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• 2008

In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor(SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.28 no.7
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• pp.1-12
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• 2014

This paper is proposed to design the new heat-sink apparatus for improving the heat transfer characteristics in the power LED chip, and results of the operation characteristics were discussed. The core design is that the soldering through-hole on the FR-4 PCB board is formed to the effective heat transfer. That is directly filled with Ag-nano materials, which shows the high thermal conductivity. The heat transfer medium consisting of Ag-nano materials is classified into two structures. Mediums are called as the heat slug and the multi-pin in this work. The heat of the high temperature generated from the LED chip was directly transferred to the heat slug of the one large size. And the accumulated heat from the heat slug was quickly dissipated by the medium of the multi-pin, which is the same body with the heat slug. This multi-pin was designed for the multi-dissipation of heat by increasing the surface areas with a little pins. Subsequently, the speed of the heat transfer with this new heat-sink apparatus is three times faster than the conventional heat-sink. Therefore, the efficiency of the illuminating light will be improved by adapting this new heat-sink apparatus in the large area's LED.