• Journal of Digital Contents Society
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• v.5 no.3
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• pp.219-224
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• 2004

In This paper, we have analyzed the power spectrum of DS-TH Ulhawideband (Direct Sequence-Time Hopping UWB) system which used pseudo-noise (PN) code. The DS-TH UWB system proposed in this paper multiplies the information signal with PN code to construct pulse train with random pattern and then the chips in pulse train are bundled into several groups to map to the particular value. The (+)/(-) pulse is tented in the time slot of frame by comparing a particular value with timing information that was stored in the lookup table. Thus, the energy spark (Comb Line) which is generated certainly in convantional system can be suppressed efficiently by PN code. And we knew that the proposed DS-TH UWB System even could have very smoothing power spectrum ctaracteristic without applying high speed Time-Hopping code.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.587-594
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• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Journal of the Korean Institute of Telematics and Electronics S
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• v.34S no.2
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• pp.1-10
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• 1997

In this paper, an improved lower bound on the burst-error correcting capability of th ecyclic product code is presented and through the analysis of this new bound clustered cyclic product (CCP abbr.)code is proposed. The CCP code, to improve the burst-error correcting capability, combines the idea of clustering and the transmission method of cyclic product code. That is, a cluster which is defined in this paper as a group of consecutive code symbols is employed as a new transmission unit to the code array transmission of cyclic product code. the burst-error correcting capability of the CCP code is improved without a loss in the random-error correcting capability and performance comparison in the digital video camera records (DVCR) system shows the superiority of the proposed CCP code over conventional product codes.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Proceedings of the KOSOMBE Conference
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• v.1992 no.05
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• pp.118-120
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• 1992

We have developed and been using laboratory order communication system which is a computerized laboratory request and reception system wi th bar code between inpatient or outpatient and the clinical laboratory in Chungbuk National Unversity Hospital. Work flows are as follows: Tests are requested by the physicians through hospital information system without issuing request forms. Bar code stickers containing demographics of patient and other informations such as sample number, slip code and specimen code are printed and attached to smaple tubes. At the department of clinical pathology, smaples are received through the bar code reader. Area numbers are automatically created and laboratory work numbers are determined. Worklists can be issued by each section of laboratory when needed. Our order communication system alleviates the human labor such as specimen labelling and making worklist and reduces clerical errors that occur from sample collection to laboratory analysis.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4373-4391
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• 2022

Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate the hydrodynamic phenomena in the reactor under motion conditions; however, its applicability does not cover the MULTID component used in multidimensional flow analyses. In this study, a moving reactor model is implemented for the MULTID component to address the importance of multidimensional flow effects under dynamic motion. The concept of the volume connection is generalized to facilitate the handling of the junction of MULTID. Further, the accuracy in calculating the pressure head between volumes is enhanced to precisely evaluate the additional body force. Finally, the Coriolis force is modeled in the momentum equations in an acceleration form. The improvements are verified with conceptual problems; the modified model shows good agreement with the analytical solutions and the computational fluid dynamic (CFD) simulation results. Moreover, a simplified gravity-driven injection is simulated, and the model is validated against a ship flooding experiment. Throughout the verifications and validations, the model showed that the modification was well implemented to determine the capability of multidimensional flow analysis under ocean conditions.

• Proceedings of KOSOMES biannual meeting
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• 2006.11a
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• pp.47-51
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• 2006

The variation of marine accidents and ship's insurances is to be the measurement of assessment concerned effect of importing ISM Code because each kind of ship accident's expenses and compensation for the loss have an effect on the ship's insurance and P&I insurance. Therefore in this study, we had grasped the accomplishment of importing ISM Code, using analysis of the variation of marine accidents and ship's insurances in domestic shipping companies from one year ago to 8 years after importing ISM Code. As the result of analysis, compared with the period of a year before carrying out ISM Code, marine accidents were showed a decrease of about 14.4% at 8th year of importing ISM Code. Also, the insurance had tended downward every year.