• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.273-280
• /
• 2014

The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software 'K-SWR'. The detectors' response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403). The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the $^{241}Am$-Be sources held in a graphite pile, a bare $^{241}Am$-Be source, and a DT neutron generator. Fluence-average energy ($E_{ave}$) varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [$H^*(10)/h$] varied from 0.99 to 16.5 mSv/h.

• Nuclear Engineering and Technology
• /
• v.56 no.8
• /
• pp.3335-3346
• /
• 2024

Due to the radiation released by commonly used isotopes, many nuclear, medical, and industrial facilities require proper radiation shielding. In this work, distinct copper borate glasses intercalated with varied aluminum and vanadium oxide (Al₂O₃ and V₂O₅) content have been synthesized and used against radiation (gamma rays and fast/thermal neutrons). The different percents were as follows: [60% B₂O₃ + 35% CuO + (5-x)% Al₂O₃ + xV₂O₅], where x = 0, 1, and 2.5 wt.%, which was coded as BCu(5-x)Al:xV. The synthesized glass samples were characterized using Fourier transforms, infrared, and X-Raydiffraction analysis. Experimentally, the radiation shielding possessions of the samples were established using an HPGe detector at the gamma energy lines 0.356 MeV, 0.661 MeV, 1.173 MeV, and 1.332 MeV. Also, the prepared glasses were investigated theoretically using the Monte Carlo code (MCNP5) at photon energies of 0.015-15 MeV. Also, the fast and thermal neutron macroscopic effective removal cross-sections were calculated using MRCsC and JANIS-4.1 software, respectively. The prepared sample BCu2.5Al:2.5V, which has a vanadium and aluminum content of 2.5%, has the highest linear attenuation coefficient as well as the highest removal cross-section for fast, and thermal neutrons.

• Environmental Mutagens and Carcinogens
• /
• v.7 no.2
• /
• pp.103-111
• /
• 1987

The effects of LET (linear energy transfer) of radiation on the induction of somatic chromosome mutations or gene mutations of Drosophila melanogaster were studied. For detecting somatic chromosome mutations and gene mutations, Drosophila wing spot system and eye-color spot system were used, respectively. The frequencies of somatic chromosome mutations or gene mutations induced after third instar larval treatment with 23 MeV neutrons, thermal neutrons, X-rays were examined. From these data, the RBE(relative biological effectiveness) values of 23 MeV neutrons relative to X-rays for induction of somatic chromosome mutations or gene mutations were calculated. The present results suggest that high LET radiations are efficient than X-ray in producing not only somatic chromosome mutations but also gene mutations.

• Journal of Radiation Protection and Research
• /
• v.26 no.2
• /
• pp.67-71
• /
• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.149-154
• /
• 1998

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute(KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons from spallation reactions are essentially required for operating the reactor in its steady state. furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance of the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases.

• Proceedings of the Korea Society for Industrial Systems Conference
• /
• 1999.05a
• /
• pp.217-222
• /
• 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
• /
• v.43 no.1
• /
• pp.75-82
• /
• 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.

• Journal of Radiation Protection and Research
• /
• v.45 no.3
• /
• pp.130-141
• /
• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3506-3513
• /
• 2023

In this paper, we demonstrate the neutron attenuation of dry cask shielding materials using the S670e helium-4 detector manufactured by Arktis Radiation Ltd. In particular, two materials expected to be applied to the TN-32 dry cask manufactured by ORANO Korea and KORAD-21 by the Korea Radioactive Waste Agency (KORAD) were utilized. The measured neutron attenuation was compared with our Monte Carlo N-Particle Transport simulation results, and the difference is given as the root mean square (RMS). For the fast neutron case, a rapid decline in neutron counts was observed as a function of increasing material thickness, exhibiting an exponential relationship. The discrepancy between the experimentally acquired data and simulation results for the fast neutron was maintained within a 2.3% RMS. In contrast, the observed thermal neutron count demonstrated an initial rise, attained a maximum value, and exhibited an exponential decline as a function of increasing thickness. In particular, the discrepancy between the measured and simulated peak locations for thermal neutrons displayed an RMS deviation of approximately 17.3-22.4%. Finally, the results suggest that a minimum thickness of 5 cm for Li-6 is necessary to achieve a sufficiently significant cross-section, effectively capturing incoming thermal neutrons within the dry cask.

• 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.