• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.30-35
• /
• 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
• /
• v.55 no.5
• /
• pp.1775-1782
• /
• 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
• /
• v.46 no.3
• /
• pp.353-362
• /
• 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
• /
• v.52 no.5
• /
• pp.869-877
• /
• 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.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
• /
• 2002.05a
• /
• pp.264-269
• /
• 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.

• Journal of Ocean Engineering and Technology
• /
• v.16 no.5
• /
• pp.1-6
• /
• 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.

• Journal of the Korea Institute of Information Security & Cryptology
• /
• v.11 no.1
• /
• pp.13-23
• /
• 2001

This paper implemented into hardware SEED which is the KOREA standard 128-bit block cipher. First, at the respect of hardware implementation, we compared and analyzed SEED with AES finalist algorithms - MARS, RC6, RIJNDAEL, SERPENT, TWOFISH, which are secret key block encryption algorithms. The encryption of SEED is faster than MARS, RC6, TWOFISH, but is as five times slow as RIJNDAEL which is the fastest. We propose a SEED hardware architecture which improves the encryption speed. We divided one round into three parts, J1 function block, J2 function block J3 function block including key mixing block, because SEED repeatedly executes the same operation 16 times, then we pipelined one round into three parts, J1 function block, J2 function block, J3 function block including key mixing block, because SEED repeatedly executes the same operation 16 times, then we pipelined it to make it more faster. G-function is implemented more easily by xoring four extended 4 byte SS-boxes. We tested it using ALTERA FPGA with Verilog HDL. If the design is synthesized with 0.5 um Samsung standard cell library, encryption of ECB and decryption of ECB, CBC, CFB, which can be pipelined would take 50 clock cycles to encrypt 384-bit plaintext, and hence we have 745.6 Mbps assuming 97.1 MHz clock frequency. Encryption of CBC, OFB, CFB and decryption of OFB, which cannot be pipelined have 258.9 Mbps under same condition.

• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.878-888
• /
• 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
• /
• v.53 no.9
• /
• pp.2788-2802
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.369-376
• /
• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.