• The Journal of Korean Institute of Communications and Information Sciences
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• v.14 no.6
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• pp.590-603
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• 1989

In tnis paper, the CRC-CCITT code and primitive polynomial CRC code are selected for analysing error detecting performance. However, general formulas for obtaining the weight distribution of these two CRC codes are not so far dericed. So, a new method for calculating the weight distribution of the shortened cyclic Hamming code is presented and an undetected error probability of these two codes is obtained when used in cell of ATM for broadband ISDN user-network interface. Consequently, we show that CRC code too much does affect its error detection performance. All the computer simulation is performed by IBM PC/AT.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1347-1354
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• 2020

To improve the efficiency of MC criticality calculation, an acceleration method of fission source convergence which gives an improved initial fission source is proposed. In this method, the MC global homogenization is carried out to obtain the macroscopic cross section of each material mesh, and then the nonlinear iterative solution of the SP3 equations is used to determine the fission source distribution. The calculated fission source is very close to the real fission source, which describes its space and energy distribution. This method is an automatic computation process and is tested by the C5G7 benchmark, the results show that this acceleration method is helpful to reduce the inactive cycles and overall running time.

• Publications of The Korean Astronomical Society
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• v.24 no.1
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• pp.43-51
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• 2009

We present a Monte-Carlo simulation code, which solves the problem of dust-scattering in interstellar dust clouds with arbitrary light source distribution and dust density structure, and calculate the surface brightness distribution. The method is very flexible and can be applied to radiative transfer problems occurring not only in a single dust cloud, but also in extragalactic dust environment. We compare, for performance test, the result of Monte-Carlo simulation with the well-known analytic approximation for a spherically symmetric homogeneous cloud. We find that the Code approximation gives a very accurate result.

• Journal of Information Processing Systems
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• v.15 no.1
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• pp.100-115
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• 2019

Malicious code distribution on the Internet is one of the most critical Internet-based threats and distribution technology has evolved to bypass detection systems. As a new defense against the detection bypass technology of malicious attackers, this study proposes the automated tracing of malicious websites in a malware distribution network (MDN). The proposed technology extracts automated links and classifies websites into malicious and normal websites based on link structure. Even if attackers use a new distribution technology, website classification is possible as long as the connections are established through automated links. The use of a real web-browser and proxy server enables an adequate response to attackers' perception of analysis environments and evasion technology and prevents analysis environments from being infected by malicious code. The validity and accuracy of the proposed method for classification are verified using 20,000 links, 10,000 each from normal and malicious websites.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• Journal of Ocean Engineering and Technology
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• v.10 no.2
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• pp.13-19
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• 1996

For fatigue analsis of offshore structures, existing methods have been reviewed and a computer code has been developed on PC. As methods to estimate the probabillity distribution of the fatigue stress, three methods(the deterministic method, the stochastic method, and the simplified method) are used in this code, to choose the appropriate method according to the situations. This code estimates damage ratios, fatigue lives, and probabilities of fatigue failure considering scatterness of SN-data, based on linear damage rule and SN-curves. Also, allowable stress for the design extreme wave can calculated by the simplified method.

• Publications of The Korean Astronomical Society
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• v.14 no.1
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• pp.23-32
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• 1999

We have developed a Monte Carlo code, which solves the problem of radiative transfer in anisotropically scattering atmosphere. The radiative code is flexible in handlings of the system geometry, the distribution of scattering particles, and the source-particle geometry. This code treats the case of highly forward throwing scattering. As performance tests, we have compared the result of Monte Carlo calculations with that of Quasi-Diffusion method for a spherically symmetric cloud model.

• Communications of Mathematical Education
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• v.6
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• pp.371-381
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• 1997

Let p be an odd prime number. We introduce simple and useful decoding algorithm for orthogonal Latin square codes of order p. Let H be the parity check matrix of orthogonal Latin square code. For any x ${\in}$ GF(p)$^{n}$, we call xH$^{T}$ the syndrome of x. This method is based on the syndrome decoding for linear codes. In L$_{p}$, we need to find the first and the second coordinates of codeword in order to correct the errored received vector.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2785-2796
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• 2023

How to generate the precise broad group cross section is important for the fast reactor design. In this study, a fast reactor multi-group cross-section generation code MGGC2.0 are developed in-house for processing ultrafine group MATXS format library. Validation and verification are performed for MGGC2.0 code by applying the benchmarks of ICSBEP handbook, and the results of MGGC2.0 agree well with that of MCNP. The consistent P_N method with critical buckling search is in good agreement that condensed with TWODANT flux and flux moment for the inner core and outer core region. For the radial blanket and reflector, two region approximation method has been applied in MGGC2.0 by using collision Probability Method neutron flux solver. The RBEC-M benchmark was used to verify the power distribution calculation, and the relative error of power distribution comparison with the reference are less than 0.8% in the fuel region and the maximum relative error is 5.58% in the reflector region. Therefore, the precise broad cross section can be generated by MGGC2.0 for fast reactor.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.