• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.689-694
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• 2013

A duplex stainless steel sheet with 1 wt.% gadolinium was fabricated for a neutron absorbing material with high strength, excellent corrosion resistance, and low cost as well as high neutron absorption capability. The microstructure of the as-cast specimen has typical duplex phases including 31% ferrite and 69% austenite. Main alloy elements like chromium (Cr), nickel (Ni), and gadolinium (Gd) are relatively uniformly distributed in the matrix. Gadolinium rich precipitates were present in the grains and at the grain boundaries. The solution treatment at $1070^{\circ}$ for 50 minutes followed by the hot-rolling above $950^{\circ}$ after keeping the sheet at $1200^{\circ}$ for 1.5 hours are important points of the optimum condition to produce a 6 mm-thick plate without cracking.

• Composites Research
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• v.36 no.6
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• pp.435-440
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• 2023

This study aimed to enhance the thermal conductivity of B₄C/Al composite materials, commonly used in transport/storage containers for spent nuclear fuel, by incorporating both boron carbide (B₄C) and cubic boron nitride(cBN) as reinforcing agents in an aluminum (Al) matrix. The composite materials were successfully manufactured through a stir casting process and practical neutron-absorbing materials were obtained by rolling the fabricated composite ingot. The evaluation of the thermal conductivity of the fabricated composites was carried out because thermal conductivity is critical for neutron absorbing materials. The thermal conductivity measurement results indicated an approximately 3% increase in thermal conductivity under the same volume fraction when compared to composite materials using only B₄C particles. Through neutron absorption cross-sectional area calculations, it was confirmed that the neutron absorption capability decreased to a negligible level. Based on the findings of this study, new design approaches for neutron absorption materials are proposed, contributing to the development of high-performance transport/storage containers.

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

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.675-680
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• 2014

Pulverization of two different sized micro-$B_4C$ particles (${\sim}10{\mu}m$ and ${\sim}150{\mu}m$) was investigated using a STS based high energy ball milling system. Shapes, generation of the impurities, and reduction of the particle size dependent on milling time and initial particle size were investigated using various analytic tools including SEM-EDX, XRD, and ICP-MS. Most of impurity was produced during the early stage of milling, and impurity content became independent on the milling time after the saturation. The degree of particle size reduction was also dependent on the initial $B_4C$ size. It was found that the STS nanoparticles produced from milling is strongly bounded with the $B_4C$ particles forming the $B_4C$/STS composite particles that can be used as a neutron absorbing nanocomposite. Based on the morphological evolution of the milled particles, a schematic pulverization model for the $B_4C$ particles was constructed.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.576-580
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• 2017

A dynamic nanoindentation method was applied to study an Fe-18 at.% Gd alloy as a neutron-absorbing material prepared by vacuum arc-melting and cast in a mold. The Fe-18 at.% Gd cast alloy had a microstructure with matrix phases and an Fe-rich primary dendrite of $Fe_9Gd$. Rietveld refinement of the X-ray spectra showed that the Fe-18 at.% Gd cast alloy consisted of 35.84 at.% $Fe_3Gd$, 6.58 at.% $Fe_5Gd$, 16.22 at.% $Fe_9Gd$, 1.87 at.% $Fe_2Gd$, and 39.49 at.% ${\beta}-Fe_{17}Gd_2$. The average nanohardness of the primary dendrite phase and the matrix phases were 8.7 GPa and 9.3 GPa, respectively. The fatigue limit of the matrix phase was approximately 37% higher than that of the primary dendrite phase. The dynamic nanoindentation method is useful for identifying local phases and for analyzing local mechanical properties.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1964-1969
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• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.525-532
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• 2016

High-energy linear accelerators are increasingly used in the medical field. However, the unwanted photo-neutrons can also be contributed to the dose delivered to the patients during their treatments. In this study, neutron fluxes were measured in a solid water phantom placed at the isocenter 1-m distance from the head of an18-MV linac using the foil activation method. The produced activities were measured with a calibrated well-type Ge detector. From the measured fluxes, the total neutron fluence was found to be $(1.17{\pm}0.06){\times}10^7n/cm^2$ per Gy at the phantom surface in a $20{\times}20cm^2$ X-ray field size. The maximum photo-neutron dose was measured to be $0.67{\pm}0.04$ mSv/Gy at $d_{max}=5cm$ depth in the phantom at isocenter. The present results are compared with those obtained for different field sizes of $10{\times}10cm^2$, $15{\times}15cm^2$, and $20{\times}20cm^2$ from 10-, 15-, and 18-MV linacs. Additionally, ambient neutron dose equivalents were determined at different locations in the room and they were found to be negligibly low. The results indicate that the photo-neutron dose at the patient position is not a negligible fraction of the therapeutic photon dose. Thus, there is a need for reduction of the contaminated neutron dose by taking some additional measures, for instance, neutron absorbing-protective materials might be used as aprons during the treatment.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.447-448
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.341.1-341.1
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• 2003

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