• Proceedings of the Korea Society for Industrial Systems Conference
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• 1999.05a
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• pp.217-222
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• 1999

In this study an objection, to be needed at public works, is to develop a measurement circuit of a gauage using radioisotope for compaction control. The developed gauage consists of measuring circuits for gamma-rays and thermal neutrons, a high voltage supply unit, and a microprocessor. To obtain meaningful numbers of pulse counts, parallel five and two circuits are provided for gamma-rays and thermal neutrons, respectively. Also, to minimize effects of natural environmental radiation and electrical noise, circuits are electrostatically shielded and pulses made by ripples are eliminated by taking frequency of high voltage supplied to the circuit and pulse height of ripples into consideration.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.360-365
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• 2020

A technique for the reduction of pulse pile-up effect in digital pulse-shape discrimination (PSD) of neutrons and gamma-rays with organic scintillation detectors is presented. The technique is based on an electronic reduction of the effective decay-time constant of scintillation pulses while retaining the PSD information of the pulses. The experimental results obtained with a NE213 liquid scintillation detector in a mixed radiation field of neutrons and gamma-rays are presented, demonstrating a figure of merit (FOM) of 1.20 ± 0.05 with an energy threshold of 350 keVee (electron equivalent energy) when the effective length of the pulses is reduced to 50 ns.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.75-82
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• 2011

The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1199-1210
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• 2017

A new in-line method for sampling neutrons emitted in (${\alpha}$,n) reactions based on alpha particle source information has been developed for continuous-energy Monte Carlo simulations. The new method uses a continuous-slowing-down model coupled with (${\alpha}$,n) cross section data to precompute the expected neutron yield over the alpha particle lifetime. This eliminates the complexity and computational cost associated with explicit charged particle transport. When combined with an integrated alpha particle decay source sampling capability, the proposed method provides an efficient and accurate method for sampling (${\alpha}$,n) neutrons based solely on nuclide inventories in the problem, with no additional user input required. Results from several example calculations show that the proposed method reproduces the (${\alpha}$,n) neutron yields and energy spectra from reference experiments and calculations.

• Nuclear Engineering and Technology
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• v.4 no.2
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• pp.102-108
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• 1972

Time-dependent neutron transport equation with delayed neutrons is analytic-ally solved in the case of isotropic scattering with constant cross sections. The equations in the two divided time regions are obtained from the original equation by the asymptotic method. It is shown that the approximate solutions in each time region are uniformly valid in time to the order of the inverse magnitude of the velocity.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4067-4071
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• 2021

The ¹³C (α,n)¹⁶O reaction cross-section is important data for nuclear physics, astrophysical, and neutrino physics experiments, however, they exhibit uncertainties due to the discrepancies in the experimental data. In this study, using the Nedis-2m program code, the energy spectrum of α-induced neutrons in a thin carbon target was calculated and the corresponding reaction cross-section was refined in the alpha particle energy range of 5-8 MeV. The results were used to calculate the intensity and energy spectrum of background neutrons produced in the liquid scintillator of KamLAND. The results will be useful in a variety of astrophysical and neutrino experiments especially those based on LS or Gd-LS detectors.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.943-948
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• 2019

In this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. $^{239}Pu$ conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1723-1729
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• 2021

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H^*(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H^*(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.160-168
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• 1983

About 25 years ago, halfway along the recorded history of the neutron as a separate entity, Korea entered the nuclear age and initiated its own neutron research and development programs. Since that time Korean scientists have taken all possible advantages of the special opportunities offered by the neutron. Scientists the world over, in the Far East, hear East, and the West, have adapted these opportunities to their special needs. These needs are manifested in all phases of modern life, including power generation by nuclear means, food preservation, production of new types of food-bearing plants, commercial uses of activation analysis, irradiations, and isotope production, nuclear medicine, industrial quality control through nuclear measurements, and direct use of neutrons in research in many areas including solid state physics, chemistry, physics, biology, and medicine. Research with neutrons has been successfully conducted using nuclear research reactors of all sizes ranging from the very small (∼10 kilowatts) to the very large(50-100 Megawatts). This speaker has teen associated with nuclear research since 1945 and directly with neutron research since 1957. From this continuous research and development activity, he will report on some of the prospects in the second 50 years of the neutron.