• Nuclear Engineering and Technology
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• 제54권1호
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• pp.30-35
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• 2022

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2625-2632
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• 2024

In this study, we investigated the possibility of employing accident tolerant fuel (ATF) in VVER-1200/V491 assembly without gadolinium-containing fuel rods using the Monte Carlo code Serpent 1.1.7 with ENDF/B-VII cross-section library. The analysis involves assembly design with reflective boundary conditions. To compare the neutronic performances, U-5Mo, U-7.5Mo, U-10Mo, and U-15Mo fuels were chosen in addition to ordinary UO₂ fuel. The concentration of ¹³⁵Xe, ¹⁴⁹Sm, fissile and fertile isotopes with burnup, reactivity feedback with fuel temperature variation, and β eff values were calculated. The results indicate that the fuel cycle length increases by 54.27% for U-5Mo, 32.6% for U-7.5Mo, and 13.8% for U-10Mo, while it decreases by 16.4% for U-15Mo fuel. Additionally, the effect of ⁹⁵Mo content in natural Mo was investigated by reducing the ⁹⁵Mo concentration. According to the results, each proposed fuel's fuel cycle length extended when the depletion ratio of ⁹⁵Mo increased. Additionally, the calculations for reactivity feedback guarantee safe operating conditions for all U-xMo fuels.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.353-362
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• 2014

The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The "Fast Total Monte Carlo" method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on $k_{\infty}$, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.869-877
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• 2020

A 24-month Advanced Power Reactor 1400 (APR1400) core with a very-low-boron (VLB) concentration has been investigated for an inherently safe and high-performance PWR in this work. To develop a high-performance APR1400 which is able to do the passive frequency control operation, VLB feature is essential. In this paper, the centrally-shielded burnable absorber (CSBA) is utilized for an efficient VLB operation in the 24-month cycle APR1400 core. This innovative design of the VLB APR1400 core includes the optimization of burnable absorber and loading pattern as well as axial cutback for a 24-month cycle operation. In addition to CSBA, an Er-doped guide thimble is also introduced for partial management of the excess reactivity and local peaking factor. To improve the neutron economy of the core, two alternative radial reflectors are adopted in this study, which are SS-304 and ZrO₂. The core reactivity and power distributions for a 2-batch equilibrium cycle are analyzed and compared for each reflector design. Numerical results show that a VLB core can be successfully designed with 24-month cycle and the cycle length is improved significantly with the alternative reflectors. The neutronic analyses are performed using the Monte Carlo Serpent code and 3-D diffusion code COREDAX-2 with the ENDF/B-VII.1.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2002년도 춘계학술대회 논문집
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• pp.264-269
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• 2002

The wave profiles of directional breaking waves are investigated experimentally in a directional wave basin. The directional breaking waves are generated by component wave focusing both in direction and frequency based on constant wave steepness and constant wave amplitude spectrum models. the profile parameters of wave crest steepness and asymmetry are adapted to analyze the evolution of breaking ware characteristics in a view of focusing efficiency. The generated breaking waves are classified into the incipient, single and multi breaking waves.

• 한국해양공학회지
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• 제16권5호
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• pp.1-6
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• 2002

Extreme waves are generated in a model basin based on directional wave focusing. The targeted wave field is described by double summation method and it is applied to serpent-type wavemaker system. The extreme crest amplitude at a designed location is obtained by syncronizing the phases and focusing the directions of wave components. Two distinguished spectrums of constant wave amplitude and constant wave steepness are adapted to describe the frequency distribution of component waves. The surface profile of generated wave packets is measured by wave guage array and the effects of dominant spectral parameters governing extreme wave characteristics are investigated. It is found that frequency bandwidth, center frequency, shape of frequency spectrum and directional range play a significant role in the wave focusing. In particular, the directional effect significantly enhances the wave focusing efficiency.

• 정보보호학회논문지
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• 제11권1호
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• pp.13-23
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• 2001

본 논문에서는 우리 나라 128 비트 블록 암호 알고리즘 표준인 SEED를 하드웨어로 구현하였다. 먼저 하드웨어 구 현 측면에서 SEED를 같은 비밀키 블록 암호 알고리즘으로 AES 최종 후보 알고리즘인 MARS, RC6, RIJNDAEL, SERPENT, TWOFISH와 비교 분석하였다. 동일한 조건하에서 분석한 결과, SEED는 MARS, RC6, TWOFISH보다는 암호 화 속도가 빨랐지만, 가장 빠른 RIJNDAEL보다는 약 5배정도 느렸다. 이에 속도 측면에서 우수한 성능을 가질 수 있는 고속 SEED 구조를 제안한다. SEED는 동일한 연산을 16번 반복 수행하므로 1라운드를 Jl 함수 블록, J2 함수 블록, key mixing 블록을 포함한 J3 함수 블록의 3단계로 나누고, 이를 파이프라인 시켜 더 빠른 처리 속도를 가지도록 하였다. G 함수는 구현의 효율성을 위해 4개의 확장된 4바이트 SS5-box 들의 xor로 처리하였다. 이를 Verilog HDL을 사용하여 ALTERA FPGA로 검증하였으며, 0.5um 삼성 스탠다드 셀 라이 브러리를 사용할 경우 파이프라인이 가능한 ECB 모드의 암호화와 ECB, CBC, CFB 모드의 복호화 시에는 384비트의 평문을 암복호화하는데 총 50클럭이 소요되어 97.1MHz의 클럭에서 745.6Mbps의 성능을 나타내었다. 파이프라인이 불 가능한 CBC, OFB, CFB 모드의 암호화와 OFB 모드의 복호화 시에는 동일 환경에서 258.9Mbps의 성능을 보였다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2788-2802
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• 2021

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.