• Journal of Radiation Protection and Research
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• v.23 no.3
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• pp.197-210
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• 1998

Regulations for the safe transport of radioactive material published by IAEA Safety Standard Series ST-l is reviewed and summarized. Safety Series No.115(International standard of radiation protection and safety for ionizing radiation and radiation sources), which reflected the new recommendation of ICRP60 published in 1991, has been a important encouragement for IAEA to revise their safety series related to the transportation of radioactive materials. IAEA Safety, Standard Series No. ST-l is summarized by comparing IAEA Safety Series No.6 regarding radiation protection system and its implementation, technical standards of packages, concept of Q system and exemption of regulation. The IAEA regulations of transportation of radioactive materials is summarized from the viewpoint of radiation protection and safety assessment. Research on transportation system of radioactive waste is suggested as a further study.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.545-552
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• 2020

The purpose of this study is to derive a lead thickness calculation formula for direct-shielded doors based on NCRP Report No.151 and IAEA Safety Report Series N0.47. After deriving the dose rate calculation formula for the direct shielded door, this formula was substituted for the lead shielding thickness calculation formula to derive the shielding thickness calculation formula at the door. The lead shielding thickness calculated from the derived direct shielded door shielding thickness calculation formula was about 6% lower than that calculated by the NCRP and IAEA secondary barrier shielding thickness calculation methods. This result is interpreted as meaning that the thickness calculation is more conservative from the NCRP and IAEA secondary barrier shielding thickness calculation methods and fits well for secondary beam shielding. In conclusion, it is thought that the formula for calculating lead shielding thickness of the direct shielded door derived in this study can be usefully used in the shield design of the door.

• The Journal of the Korea Contents Association
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• v.14 no.6
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• pp.255-261
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• 2014

The generation amount of radioactive waste has been rapidly increased by increase of the usage of radioisotope source in medical field. Especially, the use of the radioactive source of I-131 with short half-life of 8.02 days used in treatment of thyroid has been increased, and all of the wastes concerned have been disposed by means of the self-disposal method. IAEA proposed criteria for clearance level of waste which depends on the individual (10 ${\mu}Sv/y$) and collective dose (1 man-Sv/y), and concentration of each nuclide (IAEA Safety Series No 111-P-1.1, 1992 and IAEA RS-G-1.7, 2004). In this study, various radioactive wastes in medical fields are collected and measured for establishing the disposal methods and procedures of radioactive wastes. In addition, comparison evaluation of decay storage period between the half-life which was calculated by attenuation curve based on real measurement and analytical half-life is considered. With comparing the theoretical half-life and the effective half life(7.72 days) which was based on the decay equation of measured data, it is resulted in the theoretical half-life is longer than effective half-life. The storage period of radioactive waste for self-disposal may be curtailed. The result of this study will be proposed as ISO standard.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.27-35
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• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

• Journal of radiological science and technology
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• v.40 no.4
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• pp.633-638
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• 2017

The amount of radioactive waste has been rapidly increased with development of radiation treatment in medical field. Recently, it has been a common practice to use I-131 for thyroid cancer, F-18 for PET/CT and Tc-99m for diagnosis of nuclear medicine. All the wastes concerned have been disposed of by means of the self-disposal method, for example incineration, after storage enough to decay less than clearance level. IAEA proposed criteria for clearance level of waste which depends on the individual ($10{\mu}Sv/y$) and collective dose (1 man-Sv/y), and concentration of each nuclide (IAEA Safety Series No 111-P-1.1, 1992 and IAEA RS-G-1.7, 2004). In this study, specific radioactivity of radioactive waste contaminated with Re-186 was measured to confirm whether it meets the clearance level. Re-186 has long half life of 3.8 days relatively and emits beta and gamma radiation, therefore it can be applied in treatment and imaging purposes. The specific radioactivity of contaminated gloves weared by radiation workers was measured by MCA(Multi-channel Analyzer) which was calibrated by reference materials in accordance with the measuring procedure. As a result, comparison evaluation of decay storage period between the half-life which was calculated by attenuation curve based on real measurement and physical half-life was considered, and it is showed that the physical half-life is longer than induced half-life. Therefore, the storage period of radioactive waste for self-disposal may be curtailed in case of application of induced half-life. The result of this study will be proposed as ISO standard.