• IEMEK Journal of Embedded Systems and Applications
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• v.11 no.2
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• pp.97-105
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• 2016

Microcontrollers (MCUs) for endpoint smart sensor devices of internet-of-thing (IoT) are being implemented as system-on-chip (SoC) with on-chip instruction flash memory, in which user firmware is embedded. MCUs directly fetch binary code-based instructions through bit-line sense amplifier (S/A) integrated with on-chip flash memory. The S/A compares bit cell current with reference current to identify which data are programmed. The S/A in reading '0' (erased) cell data consumes a large sink current, which is greater than off-current for '1' (programmed) cell data. The main motivation of our approach is to reduce the number of accesses of erased cells by binary code level transformation. This paper proposes a built-in write/read path architecture using binary code inversion method based on hot-spot region detection of instruction code access to reduce sensing current in S/A. From the profiling result of instruction access patterns, hot-spot region of an original compiled binary code is conditionally inverted with the proposed bit-inversion techniques. The de-inversion hardware only consumes small logic current instead of analog sink current in S/A and it is integrated with the conventional S/A to restore original binary instructions. The proposed techniques are applied to the fully-custom designed MCU with ARM Cortex-M0$^{TM}$ using 0.18um Magnachip Flash-embedded CMOS process and the benefits in terms of power consumption reduction are evaluated for Dhrystone$^{TM}$ benchmark. The profiling environment of instruction code executions is implemented by extending commercial ARM KEIL$^{TM}$ MDK (MCU Development Kit) with our custom-designed access analyzer.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.91-102
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• 2015

A thermal-hydraulic design and performance analysis computer code for a once-through steam generator using straight tubes is developed. To benchmark the developed physical models and computer code, an once-through steam generator developed by other designer is simulated and the calculated results are compared with the design data. Also, the same steam generator is analyzed with the best-estimate thermal-hydraulic system code, MARS, for the code-to-code validation. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by the developed code show general agreements with the published design data as well as the analysis results of MARS. It is demonstrated that the developed code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

• Ocean Systems Engineering
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• v.9 no.3
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• pp.241-259
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• 2019

This paper presents a fully three-dimensional numerical approach for analyzing deepwater drilling riser-conductor system vortex-induced vibrations (VIV) including nonlinear soil-structure interactions (SSI). The drilling riser-conductor system is modeled as a tensioned beam with linearly distributed tension and is solved by a fully implicit discretization scheme. The fluid field around the riser-conductor system is obtained by Finite-Analytic Navier-Stokes (FANS) code, which numerically solves the unsteady Navier-Stokes equations. The SSI is considered by modeling the lateral soil resistance force according to nonlinear p-y curves. Overset grid method is adopted to mesh the fluid domain. A partitioned fluid-structure interaction (FSI) method is achieved by communication between the fluid solver and riser motion solver. A riser-conductor system VIV simulation without SSI is firstly presented and served as a benchmark case for the subsequent simulations. Two SSI models based on a nonlinear p-y curve are then applied to the VIV simulations. Also, the effects of two key soil properties on the VIV simulations of riser-conductor systems are studied.

• Journal of the Korean Society of Propulsion Engineers
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• v.21 no.6
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• pp.49-56
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• 2017

The current study has developed an in-house 3D FEM code in order to model thermoacoustic problems in an annular system and compared the acoustic field calculation results with measured ones from a benchmark combustor. From the comparison of calculation results with the measured data, the current acoustic code could successfully capture the various acoustic mode found in the annular system. In addition, it was found that the transverse waves in the combustor were strongly affected by the nozzle acoustic impedances, as well, the pressure distributions were closely related with the combustor acoustic pressure field.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.461-468
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• 2020

The improvements of the DeCART/MUSAD code system for uncertainty analysis of HTGR neutronic parameters are presented in this paper. The function for quantifying an uncertainty of critical-spectrumweighted few group cross section was implemented using the generalized adjoint B₁ equation solver. Though the changes between the infinite and critical spectra cause a considerable difference in the contribution by the graphite scattering cross section, it does not significantly affect the total uncertainty. To reduce the number of iterations of the generalized adjoint transport equation solver, the generalized adjoint B₁ solution was used as the initial value for it and the number of iterations decreased to 50%. To reflect the implicit uncertainty, the correction factor was derived with the resonance integral. Moreover, an additional correction factor for the double heterogeneity was derived with the effective cross section of the DH region and it reduces the difference from the complete uncertainty. The code system was examined with the MHTGR-350 Ex.II-2 3D core benchmark. The k_eff uncertainty for Ex.II-2a with only the fresh fuel block was similar to that of the block and the uncertainty for Ex.II-2b with the fresh fuel and the burnt fuel blocks was smaller than that of the fresh fuel block.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.618-623
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• 1998

A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between AMISN and MCNP estimates, which may require a consistent library generation for both codes.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.902-909
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• 2006

In this study, three-dimensional rotordynamic analyses have been conducted using equivalent beam, hybrid and fun three-dimensional models. The Present computational method is based on the general finite element method with rotating gyroscopic effects of a rotor system. General purpose commercial finite element code, SAMCEF which includes practical rotordynamics module with various types of rotor analysis methods and bearing elements is applied. For the purpose of numerical verification, comparison study for a benchmark rotor model with support bearings is performed first. Detailed finite element models based on three different modeling concepts are constructed and then computational analyses are conducted for the realistic and complex three-dimensional rotor system. The results for rotor stability and mass unbalance response are presented and compared with the experimental vibration test conducted in this study.