• Journal of Korea Society of Industrial Information Systems
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• v.7 no.4
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• pp.37-45
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• 2002

In this study, to measure the density and moisture of compaction, it is designed to use the 2 neutron and 5 gamma-ray detectors. Also, it can be know that Co-60 gamma-ray and Cf-252 neutron source are more useful for this system, because discharge of radiation are better than other sources. They are more profitable to improve their accuracy. Especially, to remove the interference of their reciprocal action between neutron and gamma-ray detectors it is designed shield materials with lead between 2 neutron and 5 gamma-ray detectors each other. It is performed the optimal design to shield under the suggested values completely. Because this system will use portable, so decrease of their weight are so important. For this reason, it can be decrease their weight of 5.2[kg] with shield material between each detectors and it can be think mote easy to carry and use on the industrial Spot.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.357-357
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.361.1-361.1
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• 2003

• Journal of The Korean Radiological Technologist Association
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• v.29 no.1
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• pp.86-86
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.395.1-395.1
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• 2003

• Radioisotope journal
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• v.12 no.4
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• pp.73-77
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• 1997

• The Journal of the Korea Contents Association
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• v.17 no.9
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• pp.69-76
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• 2017

This study evaluated the activation of the shielding wall and the components around the accelerator by using the medical linear accelerator. We performed simulations for energy values of 20 MV with the operating time ranging from day 1 to 30 years, and linear accelerator head and shielding wall concrete were also evaluated. The results showed that neutrons in large quantities were analyzed using high energy around thetarget point where photons were formed. Based on the activation analysis with these results, radioactivity increased with an increase in operation time and activated nuclides usually start saturating in10 years. Furthermore, the general types of nuclides formed owingto the activation were Co-60, W-181, 185, 187, Na-24, Ca-45, Mn-54, 56, and Fe-55, 59.

• Korean Journal of Materials Research
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• v.8 no.5
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• pp.457-463
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• 1998

New neutron shielding materials, KNS-201, KNS-301 and KNS-601 have been fabricated to be used for radioactive material shipping and storage cask. The base materials are a modified and a hydrogenated bisphenol- A type and novolac type epoxy resin, and aluminium hydroxide and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to form this resin shield into complicated geometric shapes such as radioactive material shipping and storage cask. Several measurements were made for the shielding materials to evaluate the thermal and mechanical properties and radiation resistance. The properties of the shielding materials are as follows: onset temperatures 2S7~28$0^{\circ}C$, thermal conductivities 0.9S~1.14W/m. K, thermal expansion coefficients 0.77~1.26x$10_{-6}{\circ}C_{-1}$, combustion characteristics < 80$0^{\circ}C$, ATB(average time of burning) < 5sec, AEB(average extent of burning) < 5mm, tensile strengths 2.5~3.2kg/$\textrm{mm}^2$, compressive strengths 13.2~1S.2kg/$\textrm{mm}^2$, flexural strengths 5.2 -6.4kg/$\textrm{mm}^2$. In general, the concerned properties of KNS-201, KNS-301 and KNS-601 were revealed to be better than those of NS-4- FR. foreign neutron shielding material. It is also observed that the radiation resistance of KNS- 601 was better than those of KNS-201 and KNS-301.

• The Journal of the Korea Contents Association
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• v.16 no.10
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• pp.758-765
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• 2016

This study analyzed for the radioactive shielding wall, which shields the medical linear accelerator. This allows to evaluate the level of waste with respect to the shield wall, which accounts for more than half of the cost of dismantling later linac facility. In addition, by analyzing the waste processing method according we discuss the way to obtain the benefits in terms of dismantling cost. Results of the simulate, the amount sufficient to screen the amount of neutron radiation occurring in the shielding wall linac was measured. And neutron activation analysis results were analyzed nuclides more than about 20. This analysis was in excess of that, $^{24}Na$, $^{45}Ca$, $^{59}Fe$ nucleus paper deregulation concentration. The value is reduced is greater the deeper the depth of the shielding wall concentration. Based on this, three specific areas (E, F, G) was estimated to be impossible to landfill or recycling. The rest area was estimated to be buried or recycled if possible more than a predetermined depth.

• 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.