According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

Decommissioning plan of nuclear facilities require the radiological characterizations and the establishment of a decommissioning process that can ensure the safety and efficiency of the decommissioning workers. By utilizing the rapidly developed ICT technology, we have developed a technology that can acquire, analyze, and deliver information from the decommissioning work area to ensure the safety of decommissioning workers, optimize the decommissioning process, and actively respond to various decommissioning situations. The established a surveillance system that monitors nuclide inventory and radiation dose distribution at dismantling work area in real time and wireless transmits data for evaluation. Developed an evaluation program based on an evaluation model for optimizing the dismantling process by linking real-time measurement information. We developed a technology that can detect the location of dismantling workers in real time using stereovision cameras and artificial intelligence technology. The developed technology can be used for safety evaluation of dismantling workers and process optimization evaluation by linking the radionuclides inventory and dose distribution in dismantling work space of decommissioning nuclear power plant in the future.

Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.

A significant amount of piping is embedded in nuclear power plants (NPPs). In decommissioning these materials must be removed and cleaned. It can then be evaluated for radioactivity content below the release level. MARSSIM presents Derived Concentration Guideline Levels (DCGLs) that meet release guidelines. Calculating DCGL requires scenarios for the placement of embedded pipe and its long-term potential location or use. Some NPPs choose to keep the embedded pipes in the building. Because others will dismantle the building and dispose of the piping in-situ, determining the disposal option for embedded piping requires the use of measurement techniques with the sensitivity and accuracy necessary to measure the level of radioactive contamination of embedded piping and meet DCGL guidelines. The main measuring detectors used in NPPs are gas counters that are remotely controlled as they move along the inside of the pipe. The Geiger-Mueller (GM) detector is a detector commonly used in the nuclear field. Typically, this GM detector used 3-detectors that cover the entire perimeter of the pipe and are positioned at 120-degrees to each other. This is called a pipe crawler. It is very insensitive to gamma and X-ray, only measures beta-emitter and does not provide nuclide identification. The second method is a method using a high-resolution gamma-ray detector. Although not yet commercialized in many places, embedded piping is a scanning method. The technique only detects gamma-emitting nuclides, but some nuclides can be identified. Gamma-ray scanning identifies the average concentration per pipe length by the detector collimator. It is considerably longer than a pipe crawler. In addition, several techniques, including direct measurement of dose rate and radiochemical analysis after scraping sampling, are used and they must be used complementary to each other to determine the source term. Expensive sampling and radiochemical analysis can be reduced if these detectors are used to measure the radioactivity profile and to perform waste classification using scaling factor. In the actual Trojan NPP, a pipe crawler detector was used to survey the activity profile in a 26 foot of an embedded pipe. These results indicate that the geometric averaging of the factors and a dispersion values for each nuclide are constant within the accuracy factors. However, in order to accurately use the scaling factor in waste classification, it must have sample representativeness. Whether the sample through smear or scraping is representative of the radionuclide mixture in the pipe. Since the concentration varies according to the thickness of the deposit and depending on the location of the junction or bend, a lot of data are needed to confirm the reliability of the nuclide mixture. In this study, the reliability of the scaling factor, sampling representativeness and concentration measurement accuracy problems for waste classification in decommissioning NPP were evaluated and various techniques for measuring radioactive contamination on the inner surface of embedded pipes were surveyed and described. In addition, the advantages and limitations of detectors used to measure radioactivity concentrations in embedded piping are described. If this is used, it is expected that it will be helpful in determining the source term of the pipe embedded in the NPPs.

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.