• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.158-167
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• 2007

It is required that the neutron dosimetry be present to monitor the reactor vessel throughout its plant life. The Ex-vessel Neutron Dosimetry Systems which consist of sensor sets, radiometric monitors, gradient chains, and support hardware have been installed for 3-Loop plants after a complete withdrawal of all six in-vessel surveillance capsules. The systems have been installed in the reactor cavity annulus in order to characterize the neutron energy spectrum over the beltline region of the reactor vessel. The installed dosimetry were withdrawn and evaluated after a irradiation during one cycle and then compared to the cycle specific neutron transport calculations. The reaction rates from the measurement and calculation were compared and the results show good agreements each other.

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.213-217
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• 2014

Medical operations and diagnosis using interventional radiology techniques have been increased. The management and monitoring of occupational radiation exposure to the staff of interventional radiology become important, specially because they stand in close proximity to the patient. The operational radiation protection quantity, Hp(10) which can be obtained from personal dosimeter do not always represent the effective dose to the staff. So, in this study, to estimate the critical organ doses to the staff of interventional radiology, Monte Carlo calculations with mathematical human phantom and dose measurements with personal dosimeters were carried out for the major interventional radiology procedures using C-arm. Results showed that the values of Hp(10) measured by personal dosimeters were higher than critical organ doses which were calculated. And the calculated dose to thyroids was much higher than those of other critical organ doses. For the proper radiation protection of the medical staff of interventional radiology, additional radiation protection for thyroids as well as for whole body shielding like wearing a lead apron should be considered.

• Nuclear Medicine and Molecular Imaging
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• v.41 no.4
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• pp.309-316
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• 2007

Purpose: Sentinel lymph node biopsy became the standard procedure in early breast cancer surgery. Faculty members might be exposed to a trace amount of radiation. The aim of this study is to quantify the radiation exposure and verify the safety of the procedure and the facilities, especially during pathologic process. Materials and Methods: Sentinel lymph node biopsies with Tc-99m human serum albumin were performed as routine clinical work. Exposed radiation doses were measured in pathologic technologist, nuclear medicine technologist, and nuclear medicine physician using a thermoluminescence dosimeter (TLD) during one month. We also measured the residual radioactivities or absorbed dose rates, the exposure distance and time during procedure, the radiation dose of the waste and the ambient equivalent dose of the pathology laboratory. Results: Actual exposed doses were 0.21 and 0.85 (uSv/study) for the whole body and hand of pathology technologist after 47 sentinel node pathologic preparations were performed. Whole body exposed doses of nuclear medicine physician and technologist were 0.2 and 2.3 (uSv/study). According to this data and the exposure threshold of the general population (1 mSv), at least 1100 studies were allowed in pathology technologist. The calculated exposed dose rates (${\mu}$ Sv/study) from residual radioactivities data were 2.47/ 22.4 ${\mu}$ Sv (whole body/hand) for the surgeon; 0.22/ 0 ${\mu}$ Sv for operation nurse. The ambient equivalent dose of the pathology laboratory was 0.02-0.03 mR/hr. The radiation dose of the waste was less than 100 Bq/g and nearly was not detected. Conclusion: Pathologic procedure relating sentinel lymph node biopsy using radioactive colloid is safe in terms of the radiation safety.(Nucl Med Mol Imaging 2007;41(4);309-316)

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.53-59
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• 2004

A variety of factors such as the pattern of intake (acute or chronic), monitoring interval and the characteristics of the radionuclides could have a significant influence on the estimates for the intake and internal dose. The relative differences of the assessed intakes based on the assumption of an acute intake to that of a chronic intake were evaluated by using the predicted bioassay quantity in the whole body or organs for an acute and chronic intake through the inhalation of $^{125}$ I, $^{137}$ C, $^{235}$ U with the AMAD of 1 ${\mu}{\textrm}{m}$ and 5 ${\mu}{\textrm}{m}$ for the monitoring intervals of 7, 14, 30, 60, 90, 120, 180, 360 days, respectively, The relative difference of the assessed intakes based on the intake pattern is affected by the monitoring interval, radionuclide and absorption type, but the particle size has little influence on the difference of the assessed intakes based on the intake pattern. The maximum monitoring interval, which is defined as the monitoring interval that the relative difference of the assessed intakes based on the assumption of an acute intake to that of a chronic intake is less than 10%, is 60 d for $^{125}$ I with Type F, 180 d for $^{137}$ C with Type F, 90 d for $^{235}$ U with Type M, and 360 d for $^{235}$ U with Type S. It was concluded that an intake pattern has little influence on the estimates of the assessed intake in the case where the monitoring interval is shorter than the maximum monitoring interval for each radionuclide.

• Journal of radiological science and technology
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• v.33 no.1
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• pp.51-59
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• 2010

We developed an analysis software that automatically detects incoming isotopes for multi-purpose gamma-ray detectors. The software is divided into three major parts; Network Interface Module (NIM), Spectrum Analysis Module (SAM), and Graphic User Interface Module (GUIM). The main part is SAM that extracts peak information of energy spectrum from the collected data through network and identifies the isotopes by comparing the peaks with pre-calibrated libraries. The proposed peak detection algorithm was utilized to construct libraries of standard isotopes with two peaks and to identify the unknown isotope with the constructed libraries. We tested the software by using GammaPro1410 detector developed by NuCare Medical Systems. The results showed that NIM performed 200K counts per seconds and the most isotopes tested were correctly recognized within 1% error range when only a single unknown isotope was used for detection test. The software is expected to be used for radiation monitoring in various applications such as hospitals, power plants, and research facilities etc.

• Journal of Radiation Protection and Research
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• v.38 no.3
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• pp.149-156
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• 2013

Recent domestic use of radiations has increased in the number of institutions and companies as well as operating as an investment, a variety of facilities and safety management are becoming increasingly complex. Despite the increase of radiation workers and facilities, the number of RSOs (Radiation Safety Officers) has not increased with a growing domestic radiation industry. The radiation safety management work (radiation workers management, radiation sources management, facilities management etc.) has been managed by insufficient number of the RSOs. These problems could be directly or indirectly related to causes of the radiation accidents. In this paper, we designed the Process Mapping of radiation safety management work for an efficient safety management of the radiation facilities and protection of radiation accidents. To develop the Process Mapping, we analyzed the radiation safety requirements of management issues and the individual procedures. Based on the Process Mapping, the work procedures for an appropriate radiation safety management of each institution can be configured clearly. Through this procedures, the safety risk factors in radiation facilities can be reduced, and the radiation safety management system will be improved. Depending on your needs, the Process Mapping could be modified and could be used for an efficient radiation safety management.