• Title/Summary/Keyword: Serpent

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A Study on MT-Serpent Cryptographic Algorithm Design for the Portable Security System (휴대용 보안시스템에 적합한 MT-Serpent 암호알고리즘 설계에 관한 연구)

  • Lee, Seon-Keun;Jeong, Woo-Yeol
    • Journal of the Korea Society of Computer and Information
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    • v.13 no.6
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    • pp.195-201
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    • 2008
  • We proposed that is suitable network environment and wire/wireless communication network, easy of implementation, security level preservation, scalable & reconfigurable to TCP/IP protocol architecture to implement suitable smart card MS-Serpent cryptographic algorithm for smart card by hardware base chip level that software base is not implement. Implemented MT-Serpent cryptosystem have 4,032 in gate counter and 406.2Mbps@2.44MHz in throughput. Implemented MS-Serpent cryptographic algorithm strengthens security vulnerability of TCP/IP protocol to do to rescue characteristic of smart card and though several kind of services are available and keep security about many user in wire/wireless environment, there is important purpose.

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Improved Result on the Pseudorandomness of SPN-type transformations (SPN 블록 암호 구조의 의사 난수성에 대한 향상된 결과)

  • 이원일
    • Journal of the Korea Institute of Information Security & Cryptology
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    • v.14 no.1
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    • pp.91-99
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    • 2004
  • Iwata et al. analyzed the pseudorandomness of the block cipher Serpent which is a SPN-type transformation. In this parer, we introduce a generalization of the results, which can be applied to any SPN-type transformation. For the purpose, we give several explicit definitions and prove our main theorems. We will also apply our theorems to several SPN-type transformations including Serpent, Crypton and Rijndael.

Improved Result on the Pseudorandomness of SPN-type transformation (SPN 구조의 의사 난수성에 대한 향상된 결과)

  • 이원일;홍석희;성재철;이상진
    • Proceedings of the Korea Institutes of Information Security and Cryptology Conference
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    • pp.57-61
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    • 2003
  • Iwata 등은 SPN 구조에 기반한 블록 암호들 중 Serpent에 대한 의사 난수성을 분석하였다 [2]. 그들은 Serpent의 구조를 최대한 보존한 상태에서 의사 난수성을 분석하기 위하여 Serpent의 Diffusion layer의 특성을 그대로 보존하여 일반화 한 후 이론을 전개하였다. 본 논문에서는 Serpent가 취한 Diffusion layer 뿐만 아니라 SPN 구조에 기반한 블록 암호들이 취할 수 있는 임의의 Diffusion layer에 대하여 적용 가능한 일반적인 이론을 도출해낼 것이다.

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Improved Differential-Linear Cryptanalysis Using DLCT (DLCT를 활용한 향상된 차분선형 분석)

  • Kim, Hyunwoo;Kim, Seonggyeom;Hong, Deukjo;Sung, Jaechul;Hong, Seokhie
    • Journal of the Korea Institute of Information Security & Cryptology
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    • v.28 no.6
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    • pp.1379-1392
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    • 2018
  • The complexity of the differential-linear cryptanalysis is strongly influenced by the probability of the differential-linear characteristic computed under the assumption of round independence, linear approximation independence, and uniformity for the trail that does not satisfy differential trail. Therefore, computing the exact probability of the differential-linear characteristic is a very important issue related to the validity of the attack. In this paper, we propose a new concept called DLCT(Differential-Linear Connectivity Table) for the differential-linear cryptanalysis. Additionally, we propose an improved probability computation technique of differential-linear characteristic by applying DLCT. By doing so, we were able to weaken linear approximation independence assumption. We reanalyzed the previous results by applying DLCT to DES and SERPENT. The probability of 7-round differential-linear characteristic of DES is $1/2+2^{-5.81}$, the probability of 9-round differential-linear characteristic of SERPENT is computed again to $1/2+2^{-57.9}$, and data complexity required for the attack is reduced by $2^{0.2}$ and $2^{2.2}$ times, respectively.

THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

  • Korkmaz, Mehmet E.;Agar, Osman
    • 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.

Karyotype Analysis of Korean Spotted Serpent Head [Channa argus (Cantor); Channiformes, Channidae] (한국산 가물치[Channa argus (Cantor);가물치목, 가물치과]의 핵형분석)

  • 이석우;이영재
    • The Korean Journal of Zoology
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    • v.29 no.2
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    • pp.75-78
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    • 1986
  • Karyotypes of Korean spotted serpent head [Channa argus (Cantor)] were analyzed to obtain a basic information on the cytogenetics of this fish. Diploid chromosome numbers were found to be 48, of which 2 were submetacentric, 10 were submeta- or subtelocentric, and 26 were acro- or telocentric chromosomes without notably hetermorphic sex chromosomes. Cytogenetical implications of the results are discussed.

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An assessment of the applicability of multigroup cross sections generated with Monte Carlo method for fast reactor analysis

  • Lin, Ching-Sheng;Yang, Won Sik
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2733-2742
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    • 2020
  • This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P0 scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P5, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC2-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in 237Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for 235U, from 2.2% to 0.7% for 238Pu, and from 1.6% to 0.2% for 241Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

Neutronic analysis of fuel assembly design in Small-PWR using uranium mononitride fully ceramic micro-encapsulated fuel using SCALE and Serpent codes

  • Hakim, Arief Rahman;Harto, Andang Widi;Agung, Alexander
    • Nuclear Engineering and Technology
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    • v.51 no.1
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    • pp.1-12
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    • 2019
  • One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

Neutronics study on small power ADS loaded with recycled inert matrix fuel for transuranic elements transmutation using Serpent code

  • Vu, Thanh Mai;Hartanto, Donny;Ha, Pham Nhu Viet
    • Nuclear Engineering and Technology
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    • v.53 no.7
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    • pp.2095-2103
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    • 2021
  • A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO2 spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library

  • Hartanto, Donny;Liem, Peng Hong
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2725-2732
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    • 2020
  • This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of 16O, 9Be, 235U, 238U, and S(𝛼,𝛽) of 1H in H2O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.