When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

Radiation workers receive exposure during radiation works such as decontamination or cutting of metals and concrete in decommissioning nuclear power plants. To reduce occupational exposure, various radiation protection measures should be prepared by estimating the exposure dose in advance. RESRAD-RECYCLE, the computer code, is generally used for estimating occupational dose due to handling metals contaminated with radioactive materials. However, RESRAD-RECYCLE used the dose conversion factors (DCF) of EPA FGR No. 11 based on ICRP Publications 30 and 48 published in the 1980s for internal exposure estimation. This study compared the DCFs of RESRAD-RECYCLE with those of the relatively recently published ICRP Publications 119 and 141. In addition, the internal exposure dose was evaluated by changing the value of the DCFs of RESRAD-RECYCLE. As a result of the comparison, ICRP Publication 119 showed that the DCF values of most nuclides were significantly lowered. On the other hand, in the case of nuclides emitting gamma rays, there was generally no significant change in the value of DCFs. In addition, in the case of 65Zn and 94Nb, the DCF increased compared to the previous ICRP publications. The exposure dose of the decommissioning workers of Hanul Units 1 and 3 and Hanbit Unit 4 was also calculated in this study. The expected radioactivity concentration of the steam generator chamber of each unit was used as the source term. The concentration of metal dust in the air generated during cutting was calculated and applied to evaluate the internal exposure dose. As a result of the dose evaluation, there was a difference in exposure dose up to 0.2 mSv in the scrap cutter scenario of Hanbit Unit 4, which generated a lot of dust and had a high radioactivity concentration. On the other hand, in the case of the slag worker, there was no difference in the dose because the working time was very short, and the inhalation of metal dust was small, even if the latest DCF was applied.