• Current Photovoltaic Research
• /
• v.6 no.1
• /
• pp.17-20
• /
• 2018

The dependency of the photovoltaic performance of p-/n-type silicon solar-cells on the metallic contaminant type (Fe, Cu, and Ni) and concentration was investigated. The minority-carrier recombination lifetime was degraded with increasing metallic contaminant concentration, however, the degradation sensitivity of recombination lifetime was lower at n-type than p-type silicon wafer, which means n-type silicon wafer have an immunity to the effect of metallic contamination. This is because heavy metal ions with positive charge have a much larger capture cross section of electron than hole, so that reaction with electrons occurs much more easily. The power conversion efficiency of n-type solar-cells was degraded by 9.73% when metallic impurities were introduced in the silicon bulk, which is lower degradation compared to p-type solar-cells (15.61% of efficiency degradation). Therefore, n-type silicon solar-cells have a potential to achieve high efficiency of the solar-cell in the future with a merit of immunity against metal contamination.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.182-189
• /
• 2018

A neutron generation system was developed to induce fissile fission in a lead slowing down spectrometer (LSDS) system. The source neutron is one of the key factors for LSDS system work. The LSDS was developed to quantify the isotopic contents of fissile materials in spent nuclear fuel and recycled fuel. The source neutron is produced at a multilayered target by the (e,${\gamma}$)(${\gamma}$,n) reaction and slowed down at the lead medium. Activation analysis of the target materials is necessary to estimate the lifetime, durability, and safety of the target system. The CINDER90 code was used for the activation analysis, and it can involve three-dimensional geometry, position dependent neutron flux, and multigroup cross-section libraries. Several sensitivity calculations for a metal target with different geometries, materials, and coolants were done to achieve a high neutron generation rate and a low activation characteristic. Based on the results of the activation analysis, tantalum was chosen as a target material due to its better activation characteristics, and helium gas was suggested as a coolant. In addition, activation in a lead medium was performed. After a distance of 55 cm from the lead surface to the neutron incidence, the neutron intensity dramatically decreased; this result indicates very low activation.

• Nuclear Engineering and Technology
• /
• v.4 no.2
• /
• pp.116-131
• /
• 1972

A fast neutron time-of-flight spectrometer has been constructed with suitable choice of target thickness and proton bombarding energy in Li$^{7}$ (p, n) Be$^{7}$ nuclear reaction for a continuous keV spectrum of neutrons at 0 degree in 1-nsec pulse from a Van do Graaff and a time-pick-up fast neutron detector assembled with a 5 mm-thick 92% enriched B$^{10}$ slab and four heavily shielded 4＂$\times$3＂ NaI scintillation detectors. Energy resolution of this spectrometer is better than 0.3% at 50 keV and the signal-to-background ratio is also improved. Total cross section measurements of several separated single isotopes have been carried out with this spectrometer and analyzed by Rmaxtrix multi-level computer code. The spin values and resonance parameters of each individual resonances are given.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3888-3896
• /
• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Journal of Radiopharmaceuticals and Molecular Probes
• /
• v.8 no.2
• /
• pp.113-118
• /
• 2022

Gadolinium neutron capture therapy (Gd-NCT) is a precision radiation therapy that kills cancer cells using the neutron capture reaction that occurs when ¹⁵⁷Gd hits thermal neutrons. ¹⁵⁷Gd has the highest thermal neutron capture cross section of 254,000 barns among stable isotopes in the periodic table. Another stable isotope, ¹⁵⁵Gd, also has a high thermal neutron trapping area (~ 60,700 barns), so gadolinium that exists in nature can be used as a Gd-NCT drug. Gd-NCT is a mixed kinetic energy of low-energy and high-energy ionizing particles, which can be uniformly distributed throughout the tumor tissue, thereby solving the disadvantage of heterogeneous dose distribution in tumor tissue. The Gd complexes of small-sized molecule are widely used as contrast agents for magnetic resonance imaging (MRI) in clinical practice. Therefore, these compounds can be used not only for diagnosis but also therapy when considering the concept of Gd-NCT. This multifunctional trial can look forward to new medical advance into NCT clinical practices. In this review, we introduce gadolinium compounds suitable for Gd-NCT and describe the necessity of image guided Gd-NCT.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.23 no.9
• /
• pp.1151-1162
• /
• 1999

The effects of radial heat and $H_2O$ diffusion on the evolution of silica particles in coflow diffusion flames have been studied experimentally. The evolution of silica aggregate particles in coflow diffusion flames has been measured experimentally using light scattering and thermophoretic sampling techniques. The measurements of scattering cross section from $90^{\circ}$ light scattering have been utilized to calculate the aggregate number density and volume fraction using with combination of measuring the particle size and morphology through the localized sampling and a TEM image analysis. Aggregate or particle number densities and volume fractions were calculated using Rayleigh-Debye-Gans and Mie theory for fractal aggregates and spherical particles, respectively. Flame temperatures and volumetric differential scattering cross sections have been measured for different flame conditions such as inert gas species, $H_2$ flow rates, and burner injection configurations to examine the relation between the formation of particles and radial $H_2O$ diffusion. The comparisons of oxidation and flame hydrolysis have also been made for various $H_2$ flow rates using $N_2$ or $O_2$ as a carrier gas. Results indicate that the role of oxidation becomes dominant as both carrier gas($O_2$) and $H_2$ flow rates increases since the radial heat diffusion precedes $H_2O$ diffusion in coflow flames used in this study. The effect of carrier gas flow rates on the evolution of silica particles have also been studied. When using $N_2$ as a carrier gas, the particle volume fraction has a maximum at a certain carrier gas flow rate and as the flow rate is further increased, the hydrolysis reaction Is delayed and the spherical particles finally evolves into fractal aggregates due to decreased flame temperature and residence time.

• Nuclear Engineering and Technology
• /
• v.46 no.3
• /
• pp.313-342
• /
• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• Nuclear Engineering and Technology
• /
• v.16 no.4
• /
• pp.202-216
• /
• 1984

The success of coarse-mesh nodal solution methods provides strong motivation for finding homogenized parameters which, when used in global nodal calculation, will reproduce exactly all average nodal reaction rates for large nodes. Two approximate theories for finding these ideal parameters, namely, simplified equivalence theory and approximate node equivalence theory, are described herein and then applied to the PWR benchmark problem. Nodal code, ANM, is used for the global calculation as well as for the homogenization calculation. From the comparative analysis, it is recommended that homogenization be carried out only for the unique type of fuel assemblies and for core boundary color-sets. The use of approximate homogenized cross-sections and approximate discontinuity factors predicts nodal powers with maximum error of 0.8% and criticality within 0.1% error relative to the fine-mesh KIDD calculations.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2725-2732
• /
• 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.

• Journal of the Korean Society of Propulsion Engineers
• /
• v.20 no.3
• /
• pp.32-36
• /
• 2016

Test equipment was designed and manufactured to evaluate thermal reaction characteristic of internal insulators of solid rocket motor. Test is allowed up to chamber pressure 2,500 psi, burn-time 100 s. A cross section of test sample part is quadrature, and various test samples can be comparable at the same time. Inner temperature of test sample can be measured by thermocouples during burning. Test was executed in condition of efficient average chamber pressure 1,000 psi, efficient burn-time 10 s and safety of equipment was confirmed. Basic data for understanding thermal characteristics of internal insulator, that is, pressure-time curve, temperature-time curve in the test sample, and thermal destruction thickness of test sample was gained successfully.