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Kim S.H.,Hanyang University | Kim J.H.,Hanyang University | Shin C.H.,Hanyang University | Kim J.K.,Hanyang University | And 2 more authors.
Nuclear Engineering and Design | Year: 2015

Vitrification form is an immobilization method for the radioactive wastes that are generated from various facilities such as nuclear power plants. Currently, the resultant vitrified forms are classified as radioactive wastes and therefore they should be disposed of in radioactive waste disposal facilities; but this disposal method is costly. In this study, a re-use method for various vitrified forms as the criticality control material was proposed for the replacement of the conventional neutron absorber in spent fuel storages. First, a nuclear fuel storage design was proposed for PLUS7 and WH 17 × 17 fuel assemblies. In addition, the criteria of the effective boron densities were established. For the verification of the criticality control ability, the multiplication factors in the storages were estimated with the 43 vitrified forms which have been developed in various countries. The results showed that most of the vitrified forms, except for a few cases, can be utilized for the criticality control materials. The proposed method uses the vitrified forms as neutron absorbers in the spent fuel storages without any chemical additions. Therefore, it can contribute to an increase in the efficient disposal of radioactive waste, as well as the providing economic benefits. © 2014 Elsevier B.V. All rights reserved. Source


Kim K.-O.,Hanyang University | Ahn W.S.,Asan Medical Center | Kwon T.-J.,Hanyang University | Kim S.Y.,RADCORE Co. | And 2 more authors.
Nuclear Engineering and Technology | Year: 2011

A sensitivity analysis of the methods used to evaluate the transport properties of a CdZnTe detector was performed using two different radiations (α particle and gamma-ray) emitted from an 241Am source. The mobility-lifetime products of the electron-hole pair in a planar CZT detector (5×5×2 mm3) were determined by fitting the peak position as a function of biased voltage data to the Hecht equation. To verify the accuracy of these products derived from α particles and low-energy gamma-rays, an energy spectrum considering the transport property of the CZT detector was simulated through a combination of the deposited energy and the charge collection efficiency at a specific position. It was found that the shaping time of the amplifier module significantly affects the determination of the (μτ) products; the α particle method was stabilized with an increase in the shaping time and was less sensitive to this change compared to when the gamma-ray method was used. In the case of the simulated energy spectrum with transport properties evaluated by the α particle method, the peak position and tail were slightly different from the measured result, whereas the energy spectrum derived from the low-energy gamma-ray was in good agreement with the experimental results. From these results, it was confirmed that low-energy gamma-rays are more useful when seeking to obtain the transport properties of carriers than α particles because the methods that use gamma-rays are less influenced by the surface condition of the CZT detector. Furthermore, the analysis system employed in this study, which was configured by a combination of Monte Carlo simulation and the Hecht model, is expected to be highly applicable to the study of the characteristics of CZT detectors. Source


Kim S.H.,Hanyang University | Park J.S.,Hanyang University | Shin C.H.,Hanyang University | Kim J.K.,Hanyang University | And 5 more authors.
Annals of Nuclear Energy | Year: 2015

In this study, a neutron absorber based on an artificial rare earth compound, which is a radioactive waste generated from pyro-process, is proposed for use in spent fuel storages. To secure the stable control of criticality with physical and chemical durability, a neutron absorber was designed and fabricated using borosilicate glass and a rare earth compound. The performance of the developed neutron absorber was evaluated in terms of the: (1) criticality controllability with various artificial rare earth compositions, (2) stability after neutron irradiation generated from the spent fuel, (3) radioactivity of the neutron absorber, and (4) physical and chemical properties. Our results show that the neutron absorber can successfully control the criticality regardless of the artificial rare earth composition. Also, we demonstrate that the neutron absorber can be utilized without any additional radiation shielding of the spent fuel storages for a long period of time (more than 100. years). In addition, analysis shows that the absorber has sufficient physical and chemical strength for use in spent fuel storage. We expect that this study will help to minimize the number of radioactive waste storage sites as well as reduce the disposal costs. © 2015 Elsevier Ltd. Source

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