• Journal of the Korean Magnetic Resonance Society
• /
• v.12 no.2
• /
• pp.74-80
• /
• 2008

The local structure of defects in undoped, Si-doped, and neutron irradiated free standing GaN bulk crystals, grown by hydride vapor phase epitaxy, has been investigated by employing electron magnetic resonance(EMR), Raman scattering and cathodoluminescence. The GaN samples were irradiated to a dose of $2{\times}10^{17}$ neutrons in an atomic reactor at Korea Atomic Energy Research Institute. There was no appreciable change in the Raman spectra for undoped GaN samples before and after neutron irradiation. However, a forbidden transition, $A_1$(TO) mode, appeared for a neutron irradiated Si-doped GaN crystal. Cathodoluminescence spectrum for the neutron irradiated Si-doped GaN crystal became much broader or was much more broadened than that for the unirradiated one. The observed EMR center with the g value of 1.952 in a neutron irradiated Si-doped GaN may be assigned to a Si-related complex donor.

• 대한방사선방어학회:학술대회논문집
• /
• 2010.04a
• /
• pp.170-171
• /
• 2010

• Journal of Radiation Protection and Research
• /
• v.37 no.2
• /
• pp.70-74
• /
• 2012

Silicon carbide (SiC) is a promising material for neutron detection at harsh environments because of its capability to withstand strong radiation fields and high temperatures. Two PIN-type SiC semiconductor neutron detectors, which can be used for nuclear power plant (NPP) applications, such as in-core reactor neutron flux monitoring and measurement, were designed and fabricated. As a preliminary test, MCNPX simulations were performed to estimate reaction probabilities with respect to neutron energies. In the experiment, I-V curves were measured to confirm the diode characteristic of the detectors, and pulse height spectra were measured for neutron responses by using a $^{252}Cf$ neutron source at KRISS (Korea Research Institute of Standards and Science), and a Tandem accelerator at KIGAM (Korea Institute of Geoscience and Mineral Resources). The neutron counts of the detector were linearly increased as the incident neutron flux got larger.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4226-4230
• /
• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Journal of Radiation Protection and Research
• /
• v.26 no.3
• /
• pp.249-253
• /
• 2001

Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of $^{24}Na$ in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of $^{24}Na$ is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 $[(Bq^{24}Na/g^{23}Na)/mGy]$.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1559-1566
• /
• 2023

This study developed a quasi-monoenergetic neutron source (QMN) for the semiconductor device's soft error rate test (SER). Quasi-monoenergetic neutrons are generated by ⁹Be(p, n)⁹B nuclear reaction with a 1 mm beryllium target and 30 MeV protons from a cyclotron. An 8 mm water in the back of the beryllium target is used for avoiding proton penetration. The neutron spectra simulated by MCNP showed that the peak energy was around 26.5 MeV. The heat flow and mechanical properties are numerically analyzed, and the safe operating conditions are therefore determined.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
• /
• v.21 no.7
• /
• pp.603-609
• /
• 2008

Photoluminescence (PL) of neutron-irradiated GaN films and nanowires is investigated in this study. The GaN films and nanowires were irradiated by neutron beams in air at room temperature, and the neutron-irradiated films and nanowires were annealed in an atmosphere of $NH_3$ at temperatures ranging from 500 to $1100^{\circ}C$. The line-shapes of the PL spectra taken from the neutron-irradiated GaN films and nanowires were changed differently with increasing annealing temperature. In this study, light-emitting centers created in the neutron-irradiated GaN films and nanowires are examined and their origins are discussed. In addition, it is suggested here that the neutron-transmutation-doping is a simple and useful means of homogeneous impurity doping into nanowires with control of the doping concentration.

• Nuclear Engineering and Technology
• /
• v.47 no.7
• /
• pp.939-944
• /
• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• Nuclear Engineering and Technology
• /
• v.56 no.6
• /
• pp.2019-2028
• /
• 2024

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

• Progress in Medical Physics
• /
• v.29 no.4
• /
• pp.101-105
• /
• 2018

In this paper, a neutron moderation system for boron neutron capture therapy (BNCT) based on a $^{252}Cf$ neutron source is proposed. Different materials have been studied in order to produce a high percentage of epithermal neutrons. A moderator with a construction mixture of $AlF_3$ and Al, three reflectors of $Al_2O_3$, BeO, graphite, and seven filters (Bi, Cu, Fe, Pb, Ti, a two-layer filter of Ti+Bi, and a two-layer filter of Ti+Pb) is considered. The MCNPX simulation code has been used to calculate the neutron and gamma flux at the output window of the neutronic system. The results show that the epithermal neutron flux is relatively high for four filters: Ti+Pb, Ti+Bi, Bi, and Ti. However, a layer of Ti cannot reduce the contribution of ${\gamma}$-rays at the output window. Although the neutron spectra filtered by the Ti+Bi and Ti+Pb overlap, a large fraction of neutrons (74.95%) has epithermal energy when the Ti+Pb is used as a filter. However, the percentages of the fast and thermal neutrons are 25% and 0.5%, respectively. The Bi layer provides a relatively low epithermal neutron flux. Moreover, an assembly configuration of 30% $AlF_3+70%$ Al moderator/$Al_2O_3$ reflector/a two-layer filter of Ti+Pb reduces the fast neutron flux at the output port much more than other assembly combinations. In comparison with a recent model suggested by Ghassoun et al., the proposed neutron moderation system provides a higher epithermal flux with a relatively low contamination of gamma rays.