• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.334-338
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• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• 한국초전도학회:학술대회논문집
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• 2008.08a
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• pp.61-61
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• 2008

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.9-14
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• 2009

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.27 no.4
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• pp.199-202
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• 2014

The effect of neutron irradiation on the properties of SiC Schottky Diode has been investigated. SiC Schottky diodes were irradiated under neutron fluences and compared to the reference samples to study the radiation-induced changes in device properties. The condition of neutron irradiation was $3.1{\times}10^{10}$ $n/cm^2$. The current density after irradiation decreased from 12.7 to 0.75 $A/cm^2$. Also, a slight positive shift (${\Delta}V_{th}$= 0.15 V) in threshold voltage from 0.53 to 0.68 V and a positive change (${\Delta}{\Phi}_B$= 0.16 eV) of barrier height from 0.89 to 1.05 eV have been observed by the neutron irradiation, which is attributed to charge damage in the interface between the metal and the SiC layer.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.165-175
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• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.34 no.6
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• pp.438-441
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• 2021

The electrical characteristics of single-crystal composite superconductors produced by a melting process were studied by neutron irradiation. In order to improve the current characteristics of the YBa₂Cu₃O_7-y superconductor, it is necessary to form an effective flux pinning center inside the superconductor. In this study, an increase in flux pinning was attempted through neutron irradiation onto YBa₂Cu₃O_7-y superconductors. The neutron irradiation was performed at 30 MeV for 500 sec, The electrical properties of the superconductors were measured in a magnetic field of 5 Tesla at 50 K using a magnetic properties measurement system (MPMS). After neutron irradiation, the critical current density of the YBa₂Cu₃O_7-y superconductor in a 1 Tesla magnetic field was 1×10⁵ A/cm². Once neutrons were irradiated at 30 MeV and 10 μA for 500 sec, the critical current density was observed to increase significantly. When neutrons are irradiated to a superconductor, micro-defects are created in the superconductor, and they act as flux pinning centers that hold the magnetic field generated when an electric current flows.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.637-643
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• 2022

Uranium and plutonium are required to be accounted in spent fuel head-end and major recovery area in pyro-process for safeguards purpose. The possibility of neutron resonance technique, as a nondestructive analysis, was simulated on isotopic fissile analysis for large scale process. Neutron resonance technique has advantage to distinguish uranium from plutonium directly in mixture. Simulation was performed on U235 and Pu239 assay in spent fuel and for scoping examination of assembly type. The resonance energies were determined for U235 and Pu239. The linearity in the neutron transmission was examined for the selected resonance energies. In addition, the limit for detection was examined by changing sample density, thickness and content for actual application. Several factors were proposed for neutron production and the moderated neutron source was simulated for effective and efficient transmission measurement. From the simulation results, neutron resonance technique is promising to analyze U235 and Pu239 for spent fuel assembly. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility on the reuse of fissile materials in the fuel cycle.

• 한국초전도학회:학술대회논문집
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• 2008.08a
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• pp.63-63
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• 2008

• 한국초전도학회:학술대회논문집
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• 2008.08a
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• pp.82-82
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• 2008

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1280-1286
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• 2023

MgF₂ as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF₂ were calculated, based on the phonon expansion model. The structural properties of MgF₂ were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF₂. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF₂ were reduced by 27.7% and 8.2%, respectively.