Radiation workers who handle radioisotopes, radioactive waste, nuclear material etc. may be contaminated with radioactive material due to inhalation, resulting in internal radiation exposure. For preventing radiation damage and monitoring the exposure of workers, KAERI operates a Body Radiation Measurement Laboratory. According to Article 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2017-77, “Regulation on Measurement and Calculation of Internal Radiation Dose,” The nuclear energy-related business operator with workers etc. shall establish and operate procedures and methods including the following Subparagraphs to secure the reliability of measurement of the internal radiation dose : operation and calibration of measuring instrument, inspection procedures, uncertainty of measurement, lower limit of detection and geometric configuration used for measurement. In accordance with the provision, Whole Body Counter utilized in the Body radiation Measurement Laboratory has periodic calibration / QA procedures to ensure reliability. This paper performed reliability validation of the measurement system of the Body Radiation Measurement Laboratory in the KAERI based on the performance criteria for radio-bioassay criteria presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The first criteria is MTL (Minimum Testing Level). ISO 28218 provides MTLs for each measurement category, type and nuclide. For reliable results, it is recommended to use calibration sources with higher radioactivity than the values given. The MTL for fission products in total body counting is 3 kBq and for the last 3 years the laboratory has been using sources of 6-7 kBq (Co-60, Cs-137 etc.). The second criteria is RMSE (Root Mean Square Error). It is a measure of total error defined as the square root of the sum of the square of the relative precision (SB) and the square of the relative bias (Br). The RMSE shall be lower than or equal to 0.25. The largest RMSE in the last 3 years is 0.12, and average value is 0.065, which meets the criteria. In this study, we verified the reliability of the radioactivity measurement system (WBC) based on the radio-bioassay standards presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The values were obtained using 3 years of calibration count data, and it was found that both MTL, RMSE for each nuclide met the standards with a large margin of error and were in good operating condition. This study can be applied to the maintenance, performance check, and reliability verification of similar in vivo radio-bioassay methods.

Kr-85 has a half-life of 10.7 years and it stays in the atmosphere for a long time. However it does not accumulate as an noble gas but only emits beta particles. Therefore its contribution to environmental radiation dose is lower than any other radionuclides. Kr-85 is one of the main fission products produced by nuclear fission reaction and artificial radionuclide that does not exist in nature. For these reasons, monitoring Kr-85 from the atmosphere is meaningful so that the nuclear-related facilities are recommended to control and regulate environmental emissions. Post Irradiation Examination Facility (PIEF) which located in KAERI is a facility that conducts various material and chemical experiments using the irradiated nuclear fuels. Therefore, various radionuclides can present in gaseous effluent including Kr-85. To prevent the environmental hazards and guarantee the radiation safety of the public, nuclear facilities are recommended to be equipped with stack radiation/radioactivity monitoring system, so that the Kr-85 concentration in gaseous effluent is controlled within the regulatory criteria. Particularly, the Kr-85 concentration of gaseous effluent is commonly monitored by the stack monitoring system connected to the process ventilation system from the hot cell. The monitoring system supply the information such as beta count rate, dose rate and flow rate, etc. Due to the concentration of Kr-85 in gaseous effluent is subject to regulatory guide lines, a systemized procedure for calculating Kr-85 concentration of the stack exhaust is necessary. Furthermore, the emission should be monitored whether it satisfies the regulatory standard or does not. This paper performed discussion on the process of calculating the concentration of Kr-85 in the gaseous effluent of PIEF stack from the monitoring system (NGM209, MGP), and the amount of Kr- 85 over the last 2 years emissions was calculated. In addition to calculating effluent rate of radioactive Kr-85, the Minimum Detectable Concentration (MDC) and Decision Threshold (SD) were calculated. As a result, the calculated Kr-85 concentration was below the SD during the entire period. It is considered that there are no environmental emissions of Kr-85.

Surface contaminants may attach to surfaces or objects in the radiation controlled area to cause radiation exposure, or spread out to the general environment by person and object exiting the radiation workplace. Accordingly, in radiological safety control, surface contamination monitoring is one of the important factors in workplace monitoring. When obtaining the measurement results for the monitoring, the results are accompanied by uncertainty since measurements contain numerous errors. Accordingly, the International Organization for Standardization (ISO) has published the ISO 7503 series which is comprehensive and detailed guidelines on the measurement and evaluation of surface contamination. ISO 7503-3 especially presents a mathematical model for the contamination measurement and provide calculation guidelines on measurement uncertainty evaluation, decision threshold and detection limit. This paper is focused on reevaluating and comparing the surface contamination monitoring method applied to radiation safety management practice and its results based on the measurement and evaluation method set by the International Organization for Standardization. The evaluation was performed in accordance with ISO 7503, and the current reporting method for measurement results was compared with the method recommended in ISO 11929 publication.

It is essential to provide a safe working environment for radiation workers. At a research reactor decommissioning site in Seoul (KRR1 & KRR2), radioactive waste drum disposal work is in progress. Before performing radiation work, it is necessary to determine the radioactivity of the waste drum to ensure safety. In this reason, we conducted a study to determine the detection efficiency of waste drums using the EXVol code. Determination of the full energy absorption peak efficiency (detection efficiency) is one of the important processes of the gamma-ray activation analysis. For the large voluminous gamma-ray sources like waste drum, the geometrical and attenuation effect should be considered. EXVol (Efficiency calculator for eXtended Voluminous source) code is a detection efficiency calculation code using the effective solid angle method. EXVol can calculate both coaxial and asymmetric structure. In addition, the introduction of a collimator made it possible to reduce the radiation intensity of a high radiation source. And it is possible to determine the precise detection efficiency according to the energy of a gamma ray at a specific position of the volume source. To verify the performance of the EXVol, a high resolution gamma spectroscopy system was constructed and measurement and analysis were performed. Measurements were performed on coaxial, asymmetric and collimated structures with standard point source, standard 1 L liquid volume source and HPGe detector. The measured results were compared with the calculation results of EXVol. The relative deviation of the measurement and calculation in the coaxial and asymmetric structures was 10%, and that of the collimation structure was 20%. Results can be available in analysis of waste drums’ radioactivity determination at a specific position.