• 한국컴퓨터정보학회논문지
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• 제13권6호
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• pp.195-201
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• 2008

이동식 시스템의 보안관련 문제점 등을 해결하기 위하여 본 논문은 네트워크 환경과 유무선 통신망에 적합하며 구현의 용이성, 비도 유지, 재수정 및 재사용 할 수 있으며 TCP/IP 프로토콜 아키텍쳐에 적합한 이동식 스마트카드용 MT-Serpent 암호알고리즘을 소프트웨어 기반이 아닌 하드웨어 기반 칩 레벨로 구현하였다. 구현된 MT-Serpent 암호시스템은 크기면에서 4,032이고 throughput은 406.2Mbps@2.44MHz를 가진다. 구현된 MT-Serpent 암호알고리즘은 스마트카드 등과 같은 이동식 시스템의 특징을 살릴 수 있도록 하기 위하여 TCP/IP 프로토콜의 보안 취약성을 보강하며 유무선 환경에서 여러 종류의 서비스가 가능하고 다수의 사용자에 대한 보안을 유지하는데 주요한 목적이 있다.

• 정보보호학회논문지
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• 제14권1호
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• pp.91-99
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• 2004

Iwata 등은 SPN 구조에 기반한 블록 암호들 중 Serpent에 대한 의사 난수성을 분석하였다. 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 도출해 낼 것이다. 또한 이러한 일반적인 이론을 Serpent, Crypton, Rijindael 등과 같은 블록 암호들에 적용한 결과를 제시할 것이다.

• 한국정보보호학회:학술대회논문집
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• 한국정보보호학회 2003년도 하계학술대회논문집
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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.

• 정보보호학회논문지
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• 제28권6호
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• pp.1379-1392
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• 2018

차분선형 분석의 복잡도는 라운드 독립성, 선형 근사식 독립성, 차분 경로를 만족하지 못하는 경로에 대한 균일성 가정 아래 계산되는 차분선형 특성의 확률에 큰 영향을 받는다. 따라서 차분선형 특성의 정확한 확률을 계산하는 것은 공격의 유효성과 관련된 매우 중요한 문제이다. 본 논문은 차분선형 분석을 위한 새로운 개념 DLCT(Differential-Linear Connectivity Table)를 제안한다. 그리고 DLCT를 적용하여 선형 근사식 독립성 가정을 완화할 수 있는 차분선형 특성의 향상된 확률 계산 방법을 제안하며, DES와 SERPENT에 적용하여 기존 분석결과를 재분석한다. DES의 7-라운드 차분선형 특성의 확률은 $1/2+2^{-5.81}$, SERPENT의 9-라운드 차분선형 특성의 확률은 $1/2+2^{-57.9}$로 다시 계산되었고 공격에 필요한 데이터 복잡도는 각각 $2^{0.2}$, $2^{2.2}$배 감소한다.

• Nuclear Engineering and Technology
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• 제46권3호
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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.

• 한국동물학회지
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• 제29권2호
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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.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3639-3647
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• 2023

This paper defines a 3D full core neutronics benchmark which is based on the NuScale small modular reactor (SMR) concept. The paper provides a detailed description of the NuScale-like core, a list of expected outputs, and a reference solution to the benchmark exercises obtained with the Monte Carlo code Serpent. The benchmark was developed in the framework of the Euratom McSAFER project and can be used for verification of computational chains dedicated to 3D full-core neutronics simulations of water cooled SMRs. The paper is supplemented with a digital data set to ease the modeling process.

• Nuclear Engineering and Technology
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• 제52권12호
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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.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3405-3424
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• 2024

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient Δρ_BU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.