• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.788-793
• /
• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.1212-1213
• /
• 2004

The plasmid was used pBR 322 and ${\varphi}X174$ RF DNA. In neutron experiment, damage of pBR 322 and ${\varphi}X174$ RF DNA were observed according to increasing concentration of BSH and neutron dose. Damage of plasmid DNA appeared obvious by increasing of BSH and neutron irradiation. In ${\gamma}-$ radiation experiment, it was carried out like above neutron experiment but damages of two plasmid appeared no differences from the control unlike neutron result.

• Nuclear Engineering and Technology
• /
• v.52 no.11
• /
• pp.2607-2612
• /
• 2020

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu-Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section Σ_R (cm^-1) of fast neutrons and total attenuation coefficient μ (cm^-1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section Σ_R (cm^-1) to 0.141 compared to 0.094 of ordinary concrete.

• Journal of Radiation Industry
• /
• v.17 no.3
• /
• pp.293-297
• /
• 2023

The burnup profile of spent fuel should be determined accurately for the safety storage of spent fuel. In this study, a neutron detection system was developed as a part of basic research to analyze the burnup profile of spent fuel, and a performance was evaluated using a radiation source. The prototype of the neutron detection system was based on a ³He proportional chamber. The ³He proportional chamber is often used for neutron measurement and analysis because of its high neutron detection efficiency and simplicity for gamma ray rejection. For quantitative evaluation, tests were conducted using calibrated ²⁵²Cf and ¹³⁷Cs sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

• Radiation Oncology Journal
• /
• v.19 no.1
• /
• pp.66-73
• /
• 2001

Purpose : For the research of Boron Neutron Capture Therapy (BNCT), fast neutrons generated from the MC-50 cyclotron with maximum energy of 34.4 MeV in Korea Cancer Center Hospital were moderated by 70 cm paraffin and then the dose characteristics were investigated. Using these results, we hope to establish the protocol about dose measurement of epi-thermal neutron, to make a basis of dose characteristic of epi-thermal neutron emitted from nuclear reactor, and to find feasibility about accelerator-based BNCT. Method and Materials : For measuring the absorbed dose and dose distribution of fast neutron beams, we used Unidos 10005 (PTW, Germany) electrometer and IC-17 (Far West, USA), IC-18, ElC-1 ion chambers manufactured by A-150 plastic and used IC-l7M ion chamber manufactured by magnesium for gamma dose. There chambers were flushed with tissue equivalent gas and argon gas and then the flow rate was S co per minute. Using Monte Carlo N-Particle (MCNP) code, transport program in mixed field with neutron, photon, electron, two dimensional dose and energy fluence distribution was calculated and there results were compared with measured results. Results : The absorbed dose of fast neutron beams was $6.47\times10^{-3}$ cGy per 1 MU at the 4 cm depth of the water phantom, which is assumed to be effective depth for BNCT. The magnitude of gamma contamination intermingled with fast neutron beams was $65.2{\pm}0.9\%$ at the same depth. In the dose distribution according to the depth of water, the neutron dose decreased linearly and the gamma dose decreased exponentially as the depth was deepened. The factor expressed energy level, $D_{20}/D_{10}$, of the total dose was 0.718. Conclusion : Through the direct measurement using the two ion chambers, which is made different wall materials, and computer calculation of isodose distribution using MCNP simulation method, we have found the dose characteristics of low fluence fast neutron beams. If the power supply and the target material, which generate high voltage and current, will be developed and gamma contamination was reduced by lead or bismuth, we think, it may be possible to accelerator-based BNCT.

• Journal of Radiation Protection and Research
• /
• v.30 no.1
• /
• pp.35-37
• /
• 2005

The radiation detector research group at KAERI has developed a high efficiency neutron detector using a Gas Electron Multiplier (GEM). The double GEM was fabricated and operated in an Ar/Isobutane mixture. For an application to a high efficiency neutron detector, $^6Li\;or\;^{10}B$ neutron converters coated on each surface of the multi GEM foils were considered. The optimized thickness of the thin film for a neutron detection was calculated with the MCNP and SRIM. The neutron efficiency was calculated by changing the chemical components of the thin film, and the thickness of the thin film. The thermalized neutrons were measured by a GEM detector with a thin neutron converter on the drift plate.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.2138-2150
• /
• 2020

We report the status of the cold neutron triple-axis spectrometer (Cold TAS) and thermal neutron triple-axis spectrometer (Thermal TAS) installed at HANARO. Cold TAS, whose specifications are standard across the world, is in the final phase of commissioning. Proper instrument operation was confirmed through a feasibility study of phonon measurements and data analyses with resolution convolution. In contrast, Thermal TAS is in the initial phase of commissioning, and improvement of the monochromator drum is now in progress from the viewpoint of radiation shielding. In addition, we report recent activities in the development of neutron basic elements, that is, film-coated Si-wafer collimators, which are promising for use in triple-axis spectroscopy, particularly in Cold TAS.

• Journal of Radiation Protection and Research
• /
• v.42 no.3
• /
• pp.141-145
• /
• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Journal of Radiation Protection and Research
• /
• v.41 no.2
• /
• pp.141-147
• /
• 2016

Background: A project for building a neutron time-of-flight (nTOF) facility is progressing. We expect that the construction will start in early 2016. Before that, a detailed simulation based on the current architectural drawings was performed to optimize the performance of our facility. Materials and Methods: Currently, several parts had been modified or changed from the original design to reflect requirements such as the layout of the electron beam line, shape of the vacuum chamber producing a neutron beam, and the underground layout of the nTOF facility. Detailed analysis for these modifications has been done with MCNP simulation. Results and Discussion: An overview of our photo-neutron source and KAERI nTOF facility were introduced. The numerical simulations for heat deposition, source term, and radiation shielding of KAERI nTOF facility were performed and the results are discussed. Conclusion: We are expecting that the construction of the KAERI nTOF facility will start in early 2016, and these results will be used as basic data.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2075-2083
• /
• 2021

The article provides a review of the research results obtained during of more than 20 years concerning using the gaseous radiochemical method (GRCM) for detecting of ionizing radiation. This method based on threshold nuclear reactions with production of radioactive noble gas which does not interact with the materials of gaseous tract. The applications of GRCM in the diagnostics of neutrinos, neutrons, charged particles, thermonuclear plasma thermometry, and the study of the structure and dynamics of astrophysical objects, position-sensitive dosimetry of neutron targets with accelerator driving, spatial distribution of the fast neutron flux density in a nuclear reactor allowing the transformation of longitudinal coordinate of neutron flux distribution into a temporal distribution of the radiochemical gas decay counting rate ("barcode" semblance) and measurement of bombarding particles spectra are described. Experimental testing of the described technologies was made on the neutron target driven with the linear proton accelerator of Institute for Nuclear Research of Russian Academy of Sciences (INR RAS).