To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y−1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

n Korea, the decommissioning of nuclear power plants is being prepared, and a large amount of radioactive waste is expected to be generated. In particular, clearance level waste, which accounts for more than 90%, requires prevention of cross-contamination and prompt classification. In this study, the possible exposure route and the derivation of exposure dose for worker exposure management in a movable analysis system that can be analyzed onsite were studied. The movable radionuclide analysis system is divided into a preparatory room, a sample storage room, a radioanalysis room, a laboratory, and a waste storage room. It consists of one radioanalysis worker and one pre-treatment worker, and the main radiation exposure is expected to occur in the movement path in the sample storage room, radioanalysis room, and laboratory. The source term for the exposure evaluation, the annual usage dose presented in the radiation safety report in the movable radionuclide analysis system was used. The input data for the evaluation of the external exposure dose under normal circumstances (exposure situation, working hours, distance, etc.) is referenced at facility specifications. The internal exposure dose evaluation was assumed to be acute exposure (1 hour) assumed as internal pollution due to the drop in liquid sample during the pretreatment work. As an evaluation method, a method using a calculation formula and a method using an evaluation code was performed. For the evaluation of exposure dose using the calculation formula, a preliminary evaluation was performed using the point source method, the point kernel method, and intake and dose conversion factors. In addition, VISIPLAN and IMBA codes were used to evaluate exposure dose using the evaluation code, and the input data were supplemented for evaluation. As a result of the evaluation, the annual exposure dose limit of 20 mSv was satisfied for both normal and non-normal situations. In future research, it is planned to derive the evaluation results by particular scenarios for the detailed movement route and evaluation time according to the work process in the mobile radionuclide analysis.

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.

In the dismantling process of a reactor coolant system (RCS) piping, a radiation protection plan should be established to minimize the radiation exposure doses of dismantling workers. Hence, it is necessary to estimate the individual effective dose in the RCS piping dismantling process when decommissioning a nuclear power plant. In this study, the radiation exposure doses of the dismantling workers at different positions was estimated using the MicroShield dose assessment program based on the NUREG/CR-1595 report. The individual effective dose, which is the sum of the effective dose to each tissue considering the working time, was used to estimate the radiation exposure dose. The estimations of the simulation results for all RCS piping dismantling tasks satisfied the dose limits prescribed by the ICRP-60 report. In dismantling the RCS piping of the Kori-1 or Wolsong-1 units in South Korea, the estimation and reduction method for the radiation exposure dose, and the simulated results of this study can be used to implement the radiation safety for optimal dismantling by providing information on the radiation exposure doses of the dismantling workers.

한국원자력환경공단은 처분시설 내 1단계 인수·저장구역의 인수검사 공간 및 드럼 취급 공간 부족에 대한 문제를 해결하기 위하여 방폐물검사건물을 건설하여 저장·처리능력을 확충할 예정이다. 본 연구에서는 MCNP 코드를 이용하여 방폐물검사 건물 내 저장구역에서 취급하는 해체 방사성폐기물 대상 신형처분용기를 대상으로 작업종사자의 피폭선량을 평가하였다. 평가결과, 시설 내 저장 가능한 최대 용기 개수(304개)와 방사선작업에 대한 연간 예상 작업시간(약 306시간)에 대하여 연간 집단선량은 총 84.8 man-mSv로 계산되었다. 시설 내 총 304개의 신형처분용기(소형/중형 타입)가 저장 완료된 시점에서 인수검사, 처분검사를 위한 작업종사자의 투입인력은 총 25명, 작업종사자 당 예상피폭선량은 연평균 3.39 mSv로 산출 되었다. 소형용기 취급 시 작업종사자의 고방사선량 작업에 따른 작업효율과 방사선적 안전성 확보를 위해서는 콘크리트 라이너의 두께를 증가시키는 추가적인 차폐가 필요할 것으로 평가되었다. 향후 본 연구를 바탕으로 실측기반의 해체폐기 물의 선원항과 특성을 활용하여 방사선작업 당 작업시간 및 투입인력을 산출함으로써 작업종사자의 최적의 방사선작업조건을 도출할 수 있을 것으로 사료된다.