• Journal of Welding and Joining
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• v.28 no.5
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• pp.69-74
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• 2010

Irradiation behavior of the reactor pressure vessel SA508 class 3 steel weld metals was examined by Charpy V Notch impact specimens. The specimens were exposed to a fluence of $2.8{\times}1019$ neutrons(n)/$cm^2$(E>1 MeV) at $288^{\circ}C$. The irradiation damage of weld metal was evaluated by comparison between unirradiated and irradiated specimens in terms of absorbed energy and lateral expansion. The specimens for neutron irradiation were welded by submerged arc welding process at a heat input of 3.2 kJ/mm which showed good toughness in terms of weld microstructure, absorbed energy and lateral expansion. The post-irradiation Charpy V notch 41J and 68J transition temperature elevation were $65^{\circ}C$ and $70^{\circ}C$, respectively. This elevation was accompanied by a 20% reduction in Charpy V notch upper shelf energy level. The lateral expansion at 0.9mm irradiated Charpy specimens showed temperature elevation of $65^{\circ}C$ and was greatly decreased due to radiation damage.

• Journal of the Korean Magnetics Society
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• v.15 no.1
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• pp.12-16
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• 2005

The $Fe_{87}Zr_{7}B_{6}$ amorphous alloy after neutron irradiation are studied hysteresis loop and complex permeability measurements. The total integration fluence of fast neutrons is varied from $1.92{\times}10^{14}$ to $4.85{\times}10^{16}n_{f1}cm^{-2}$. After neutron irradiation, the imaginary part of complex permeability in low frequency region decreased due to the decrease of wall motion, but the permeability in high frequency region increased due to the enhancement of rotational magnetization. The measurement of hysteresis loop showed the increase of magnetic softness, related to rotational magnetization, but saturation magnetization was decreased in neutron irradiation sample.

• Journal of the Korean Society of Radiology
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• v.1 no.1
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• pp.45-49
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• 2007

keV-neutron capture gamma-ray spectrum of $^{197}Au$(natural gold) sample have been measured in neutron energy range from 10 to 90 keV using the 3-MV pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo Institute of Technology. Pulsed keV neutrons were produced from the $^7Li(p,n)^7Be$ reaction by bombarding on the $^7Li$ target with the 1.5-ns bunched proton beam. The incident neutron spectrum on the Au sample was measured by a $^6Li$-glass scintillation detector and TOF method. Capture gamma-rays from Au sample were measured by anti-Compton NaI(TI) spectrometer. Five average neutron energy regions were selected to obtain the neutron capture spectrum. Several gamma-ray peaks in the spectrum were found in the present experiment.

• Journal of the Korean Society of Radiology
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• v.10 no.7
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• pp.489-494
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• 2016

Nuclear reaction which occurs in the cyclotron generate unnecessary neutrons. The results of this happening can radioactivate surrounding materials and radioactive materials cause radiation exposure. When people take radioactive air, it makes internal exposure. The purpose of this study was to analyze the radioactive air inside of the ultra-compact 16.5 MeV cyclotron in operation. As a result of study, the radio activation occurred by compact cyclotron generates a very low internal exposure to workers. Comparing the radioactivity from radioactive nuclide with legal standard, that was under reference value. However, it could be at risk for internal exposure in case of higher energy cyclotron. Therefore, legal standard is needed for ventilation equipment of radiation facilities.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.614-620
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• 2003

KAERI has developed DUPIC nuclear fuel with the refabrication of spent PWR fuel discharged from domestic nuclear power plant by a dry process at M6 hot-cell in IMEF To verify the performance of DUPIC nuclear fuel, irradiation test at the operating conditions of commercial power plant is essential. Since the HANARO research reactor of KAERI does not have fuel test loop(FTL) for irradiating nuclear fuel under high temperature and high pressure conditions, DUPIC fuel cannot be irradiated in the FTL of HANARO. In the 13-th PRM among Korea, Canada, USA and IAEA, AECL proposed that KAERI fabricated DUPIC fuel can be irradiated in the FTL of the NRU research reactor without charge of neutrons. The transportation quantity of DUPIC fuel to Canada is 10 elements(about 6kg). This transportation package is classified as the 7-th class according to "recommendation on the transport of dangerous goods" made by the United Nations. In case of air shipment, until now, there is no proper air transportation cask for DUPIC fuel. In case of sea transportation is possible but requires very high cost.high cost.

• Korean Journal of Crystallography
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• v.11 no.3
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• pp.162-166
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• 2000

KD₂PO₄ single crystals were grown from D₂O with reagent KH₂PO₄ and the crystal structure was determined by X-ray and neutron diffraction methods. The crystals are tetragonal at room temperature, I42d, with lattice parameters of a=7.4633(7), c=6.9785(5) Å and Z=4. Intensity data were collected on an Enraf-nonius CAD4 diffractometer with a graphite monochromated MoK/sub α/ radiation (λ=0.7107Å) and on the neutron four circle single crystal diffractometer with Ge(331) monochromated neutron beam (λ=0.997Å). The structure was refined by full-matrix least-square to final R and wR values of 0.030 and 0.072, respectively, for 204 observed reflections with I>2σ(I) by X-ray diffraction and to final R=0.041 and wR=0.096 for 144 observed relfecdtions by neutron diffraction. The O…O distance of 2.516(4)Å obtained by X-ray diffraction is the same as that of 2.515(4)Å by neutron diffraction. On the other hand, the O-D/H distance of 0.84(4)Å by X-ray diffraction is considerably shorter than 1.029(7) Åby neutron diffraction. Hydrogen and deuterium can be readily distinguished by neutrons. In this crystal 66% of H-positions were substituted by D and the rest 34% occupied by H. The phase transition temperature of DKDP obtained with deuteration levels is f193K. This value agrees fairly well with the result of DSC measurement. The nuclear density distribution by neutron diffraction provides an observation of the disordered state of D/H in KD₂PO₄ at room temperature.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.531-544
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• 2009

This paper is intended to provide key issues and current research outcomes on accelerator-based Boron Neutron Capture Therapy (BNCT). Accelerator-based neutron sources are efficient to provide epithermal neutron beams for BNCT; hence, much research, worldwide, has focused on the development of components crucial for its realization: neutron-producing targets and cooling equipment, beam-shaping assemblies, and treatment planning systems. Proton beams of 2.5 MeV incident on lithium target results in high yield of neutrons at relatively low energies. Cooling equipment based on submerged jet impingement and micro-channels provide for viable heat removal options. Insofar as beam-shaping assemblies are concerned, moderators containing fluorine or magnesium have the best performance in terms of neutron accumulation in the epithermal energy range during the slowing-down from the high energies. NCT_Plan and SERA systems, which are popular dose distribution analysis tools for BNCT, contain all the required features (i.e., image reconstruction, dose calculations, etc.). However, detailed studies of these systems remain to be done for accurate dose evaluation. Advanced research centered on accelerator-based BNCT is active in Korea as evidenced by the latest research at Hanyang University. There, a new target system and a beam-shaping assembly have been constructed. The performance of these components has been evaluated through comparisons of experimental measurements with simulations. In addition, a new patient-specific treatment planning system, BTPS, has been developed to calculate the deposited dose and radiation flux in human tissue. It is based on MCNPX, and it facilitates BNCT efficient planning based via a user-friendly Graphical User Interface (GUI).

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.