• Education of Primary School Mathematics
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• v.22 no.2
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• pp.129-147
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• 2019

Because of the various concepts and meanings of fractions and the difficulty of learning, studies to improve the teaching methods of fraction have been carried out. Particularly, because there are various methods of teaching depending on the type of fractions or the models or methods used for problem solving in fraction operations, many researches have been implemented. In this study, I analyzed the fractional operations of CCSSM-CA and its U.S. textbooks. It was CCSSM-CA revised and presented in California and the textbooks of Houghton Mifflin Harcourt Publishing Co., which reflect the content and direction of CCSSM-CA. As a result of the analysis, although the grades presented in CCSSM-CA and Korean textbooks were consistent in the addition and subtraction of fractions, there are the features of expressing fractions by the sum of fractions with the same denominator or unit fraction and the evaluation of the appropriateness of the answer. In the multiplication and division of fractions, there is a difference in the presentation according to the grades. There are the features of the comparison the results of products based on the number of factor, presenting the division including the unit fractions at first, and suggesting the solving of division problems using various ways.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.523-531
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• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3516-3525
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• 2022

The high capture cross-section (𝜎_c) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high 𝜎_c as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd₂O₃ and Er₂O₃ respectively. Models comprise nuclear fuel with a homogeneous mixture of Er₂O₃, AmO₂, SmO₂, and HfO₂ with Gd₂O₃ as well as the coating of PaO₂ and ZrB₂ on the Gd₂O₃ pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er₂O₃ and SmO₂ showed expected results namely higher values of these parameters than the reference design of Gd₂O₃ at the early burnup period. The highest value was found for the model of PaO₂ and Gd₂O₃ followed by ZrB₂ and HfO₂. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd₂O₃ and absorber coating thickness.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4287-4294
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• 2023

In order to meet the energy demand, energy production must be done continuously. Hydrogen seems to be the best alternative for this energy production, because it is both an environmentally friendly and renewable energy source. In this study, the hydrogen fuel production of the peaceful nuclear explosives (PACER) fusion blanket as the energy source integrated with Fe-Cl thermochemical water splitting cycle have been investigated. Firstly, neutronic analyzes of the PACER fusion blanket were performed. Necessary neutronic studies were performed in the Monte Carlo calculation method. Molten salt fuel has been considered mole-fractions of heavy metal salt (ThF₄, UF₄ and ThF₄+UF₄) by 2, 6 and 12 mol. % with Flibe as the main constituent. Secondly, potential of the hydrogen fuel production as a result of the neutronic evaluations of the PACER fusion blanket integrated with Fe-Cl thermochemical cycle have been performed. In these calculations, tritium breeding (TBR), energy multiplication factor (M), thermal power ratio (1 - 𝜓), total thermal power (P_hpf) and mass flow rate of hydrogen (ṁ_H2) have been computed. As a results, the amount of the hydrogen production (ṁ_H2) have been obtained in the range of 232.24x10⁶ kg/year and 345.79 x10⁶ kg/year for the all mole-fractions of heavy metal salts using in the blanket.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.23 no.3
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• pp.445-457
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• 2013

This paper is about the study to build a quantitative methodology to assess cyber threats and vulnerabilities on control systems. The SCADA system in power industry is one of the most representative and biggest control systems. The SCADA system was originally a local system but it has been extended to wide area as both ICT and power system technologies evolve. Smart Grid is a concept to integrate energy and IT systems, and therefore the existing cyber threats might be infectious to the power system in the integration process. Power system is operated on a real time basis and this could make the power system more vulnerable to the cyber threats. It is a unique characteristic of power systems different from ICT systems. For example, availability is the most critical factor while confidentiality is the one from the CIA triad of IT security. In this context, it is needed to reflect the different characteristics to assess cyber security risks in power systems. Generally, the risk(R) is defined as the multiplication of threat(T), vulnerability(V), and asset(A). This formula is also used for the quantification of the risk, and a conceptual methodology is proposed for the objective in this study.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
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• v.8 no.2
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• pp.161-172
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• 1997

In this paper, electromagnetic scattering by a perfectly conducting strip grating on dielectric multilayers is analyzed for the normalized reflected and transmitted power by applying the Fourier-Galeakin moment method. The induced current density is expanded in a series of multiplication of chebyshev polynomials of the first kind and functions with appropriate edge boundary condition, the continuous condition of electromagnetic field is applied in the boundary planes. The confirm the validity of the proposed method, the nor- malized reflected and transmitted power obtained by varying the relative permittivity and thickness of each dielectric layers are evaluated and compared with those of the existing numerical method and a paper, and then the numerical results in this paper are in good agreement with those of the existing numerical method and the paper. The sharp variation position in the geometrically normalized reflected and transmitted power can be moved by the incident angle, grating period, and the relative permittivity and thickness of the dielectric multilayers, these sharp variation points which are called the Wood's anomaly of the Geome- trically normalized reflected power are observed as a main factor when the reflected powers of the higher order mode are transitted between propagating and evanescent modes, and the local minimum positions are slightly moved to the left hand direction in which grating period is getting small according to the increase of the relative permittivity of dielectric layers.