Important medical radionuclides for Positron Emission Tomography (PET) are producing using cyclotrons. There are about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30-40 years) and actual operating duration (about 20-30 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating accelerated charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Generally, experienced cyclotron personnel prefer an unshielded cyclotron because of the repair and maintenance time. In unshielded cyclotron, water cooling systems, air compressor, and other equipment and structures could be existed for operating purposes. Almost all the equipment and structures are consisted of steel, and these affect neutron distribution in vault especially thermal neutron on the concrete wall. In addition, most of them can be classified as very low level radioactive wastes by Nuclear Safety and Security notice (NSSC Notice No. 2020-6). However, few studies were estimating radioactivity concentrations (Bq/g) of surrounding structures using mathematical calculation/simulation codes, and they were not evaluating the effect of surrounding structures on neutron distribution. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we evaluated effect of the interaction between surrounding structures (including surrounding equipment) and secondary neutrons. Discrepancies of activation distribution on/in concrete wall will be occur depending on thickness of structure, distance between structures and walls, and consideration of interaction between structures and neutrons. Throughout this study, we could find that the influence of those structures can affect neutron distribution in concrete walls even if, thickness of the structure was small. For estimating activation distribution in unshielded cyclotron vault more precisely, not only considering cyclotron components and geometry of target, but also, considering surrounding structures will be much more helpful.

As the amount of on-site Spent Nuclear Fuel (SNF) in storage increases due to the continued operation of Nuclear Power Plants (NPPs) in Korea, the on-site wet storage pool is expected to become saturated. Therefore, a facility for safely storing the spent nuclear fuel is required so that there is no problem with operation of the NPP until permanent disposal of SNF. Prior to the construction of such a facility, the safety analysis of the interim storage facility and verification of the safety of the spent fuel storage system (e.g. cask, silo) to be used are required according to Article 63 of the Nuclear Safety Act. In this process, analysis of the Structures, Systems, and Components (SSCs) of the storage system is needed. Based on the analysis, it is necessary to efficiently classify SSCs that are important to safety in order to differentiate management that more thoroughly manages those important to safety. In Korea, according to the notice of the Nuclear Safety and Security Commission, the components performing essential safety functions for the safe storage of spent fuel storage system are to be classified as “important safety equipment”. 10 CFR Part 72, a federal regulation related to interim storage facilities in the United States, also requires the identification of SSCs that fall under “Important to Safety (ITS)”, which is like domestic case. In addition, it has been confirmed that there are cases in which detailed classification according to Reg Guide 7.10 and NUREG-CR/6407 is added in Safety Analysis Report. However, these existing classification methods are not only classified as a single grade except for the method according to the Reg guide, but all are classified according to a qualitative standard. Qualitative criteria may cause ambiguity in judgment, resulting in subjective judgment of the person who proceeds in the classification process. Therefore, in this study, a new classification method is proposed to solve the problem according to the qualitative classification method. Assessing the level of radiological harm to the general public due to the assumption of failure of SSC in the spent fuel storage system is used as a quantitative evaluation standard.

Radioactive effluent discharged from the nuclear power plant (NPP) during normal operation is controlled by the discharge limit in terms of radioactivity concentration (Bq·m−3) and dose constraints in Korea. To ensure compliance with discharge limits of effluents, the licensee operates radioactive effluent monitoring systems in each discharge point to detect radioactivity and control discharge. The predetermined regulatory requirements of analytical sensitivities for sampling devices in the monitoring system are established in various countries to guarantee the performance of the monitoring systems. In Korea, Lower Limits of Detection (LLD) are selected as the regulatory requirements and adopted from the United States Nuclear Regulatory Commission (USNRC) NUREG-1301. The International Atomic Energy Agency stated that the detection limits have to be low enough (e.g., less than 1% of discharge limits) to safely demonstrate compliance with the discharge limits. However, no technical background of LLD has been explained regarding the compliance with discharge limits in Korea. Thus, it is necessary to analyze the compatibility of discharge limits and detection limits. The USNRC Regulatory Guide 1.21 has stated the risk-informed approach for effluent control by identifying the principal radionuclides whose radiological impact is more than 1% of discharge limits. In 2017, Cheong proposed the methodology and derived risk-based detection limits for liquid effluents from Korean NPPs. In 2019, Choi derived risk-based detection limits for liquid and gaseous effluents based on APR 1400 Design Control Document (DCD). The methodology of those studies can derive the detection limit for each principal radionuclide that is comparable to 1% to 10% of discharge limits. However, the previous study based on APR 1400 DCD was for the discharge limits of the US and didn’t consider the multiple discharge points in the reactor. Therefore, this study preliminarily derived the risk-based detection limits consistent with Korean Effluent Concentration Limits for gaseous effluents reflecting the characteristics of each discharge point. Also, this study confirmed the validity of risk-based detection limits and current LLD. This study is expected to be basic research for detection limits of Korean NPPs in line with international safety standards.

For producing radionuclides which were mostly used in medical purposes, for instance, Positron Emission Tomography (PET), there were about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30–40 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F, 14N(d,n)15O etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Most of the ancillary systems including peripheral equipment can be neutron activated, since, most of them were made of steels. Steels like stainless steel or carbon steel may contain some impurities, typically cobalt. Although, there were several researches evaluating activation of concrete walls and accelerator components, estimating the activation and influence on neutron interaction of the other surrounding infrastructures were insufficient. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we estimated neutron distribution in cyclotron vault and activation of ancillary subsystems including some peripheral equipment. Also, using Au foil and Cd cover, we measured thermal neutron distribution at 16 points on the concrete wall, and compared it to calculated results (MCNP). Even though, the compared results matches well, there was a discrepancy of neutron distributions between presence and absence of those equipment. Additionally, in estimating activation distributions by calculating, most of the steel-based subsystems including peripheral equipment should be managed by radioactive wastes after 20 years of operation. Throughout this study, we could find that influence on neutron interaction of those equipment can affect neutron distribution in concrete walls. This results vary the activation depth as well as location of the hot contaminated spot in concrete wall. For estimating or evaluating activation distributions in cyclotron facilities, there was need to consider some equipment located in cyclotron vault.

In worldwide, tens of thousands of units of particle accelerators have been used and more than 97% of those accelerators are used for dedicated medical of commercial applications. Radionuclide production cyclotron produce several positron-emitting radionuclides such as 18F by 18O(p,n)18F reaction which generates secondary neutrons. It is of note that these neutrons cause neutron activation in structures and components of cyclotron facilities. Therefore, International Atomic Energy Agency had addressed that a well-developed estimate of the neutron activation induced radioactive inventory of accelerator facilities is needed for the proper planning and safe implementation of decommissioning using proven methods or codes that can be used to perform activation calculations. Moreover, IAEA suggested that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be fully characterized as part of early decommissioning planning. In this study, Neutron activation in the medical cyclotron facilities was evaluated with the MCNP and FISPACT-II code to analyze the generation of decommissioning radioactive wastes during facilities dismantling. For the reference case, residual radioactivity concentration of each activation product (e.g. 60Co, 152Eu, etc.) was calculated and the sum of fractions of the activity concentration of each radionuclide divided by its clearance level was exceeded 1.0 at each calculation point which means radioactive waste generations during decommissioning of the facility. Several points show the calculated sum of fractions (SoF) at inside wall were bigger than the surface wall. The reason of these phenomena is that the slowdown of the incident neutron energy at the inside wall due to neutron attenuation and larger thermal neutron flux than surface wall. It is of note that each activation reaction cross-section was dominant at thermal neutron energy band. Sensitivity analysis was conducted to analyze the effects of design characteristics (e.g. beam energy and current, operation period, and workload). The SoF was exceeded 1.0 at the least activation condition (i.e. 9 MeV, 10 μA) if the operation period was 10 years. For the realistic condition such as 13 MeV, only 10 μA of beam current case shows the SoF was under union. On the other hand, 19 MeV, 60 μA, and 10 years operation case shows the SoF as 20.4 which means the clearance rule can be applied only after 21 years of decay-in-storage. The result of this study can be used for proper planning of decommissioning and/or new installation of cyclotron facilities include considerations of radioactive waste management.

The International Atomic Energy Agency (IAEA) refers to the possibility of changes in the discharge characteristics of radioactive effluents that are different from those during operation when a nuclear power plants (NPPs) are decommissioned. In addition, the IAEA recommends differentiated radioactive effluent management for each phase during decommissioning that reflects changes in discharge characteristics, and changes to authorization and program that are different from those in operation. Bonavigo et., al. estimated the discharge and dose of liquid and gaseous radioactive effluents based on the decommissioning plan of the Trino NPP in Italy during decommissioning, but there is a fundamental limitation in that actual data were not used. Kang and Cheong analyzed the discharge characteristics of radioactive effluents at each activities of decommissioning after permanent shutdown using actual data on radioactive effluents from the United States and Europe, and performed theoretical modeling of discharge characteristics during permanent shutdown. However, there are limitations in that only the emitted radioactivity was considered, the dose assessment was not taken into account, and the improvement methods for the differentiated monitoring program for each phase of decommissioning mentioned in the IAEA were not proposed. Most studies of radioactive effluents discharge from NPPs focus on normal operation, and studies of shutdown or decommissioned NPPs is very limited. Existing studies have not been extended to research on decommissioned NPPs, and there are limitations in that they do not consider the characteristics of decommissioned NPPs mentioned in the IAEA. Therefore, this study aims to improve the effluent monitoring program based on the analysis of the discharge characteristics NPPs that are permanently or long-term shutdown and the change in offsite dose to public. For this purpose, research was conducted on Kori Unit 1 and Wolsong Unit 1 in Korea, which were virtually permanently shutdown, and other long-term shutdown NPPs due to prolonged planned outage maintenance or replacement/repair of equipment in nuclear facility. The discharge characteristics of each radionuclide group, and further, the effect of radioactive effluent released to the environment on the offsite dose are analyzed in details.

Kori and Wolsong unit 1 were permanently shutdown in 2017 and 2019, respectively. During the decommissioning of a nuclear power plant, various types and levels of decommissioning waste will be generated sporadically in many areas in a relatively short period of time, so safe management of decommissioning waste is expected to emerge as a very important issue in the future. Since Korea has no experience in decommissioning nuclear power plants, radionuclides added by abnormal routes or errors in data can be identified through the list of expected nuclides and radioactivity data during decommissioning by analyzing cases of overseas nuclear power plants decommissioning. Therefore, it is expected that safety information of nuclear power plants in the United States (i.e. all information related to safety, such as radioactive waste characteristics and accident or decommissioning information at nuclear power plants) can be utilized when decommissioning Korea nuclear power plants. Therefore, in this study, the characteristics of solid radioactive waste were analyzed by collecting solid radioactive waste data during operation and after permanent shutdown of nine PWR nuclear power plants in the United States, and the correlation between the characteristics data of solid radioactive waste was analyzed. However, in the case of Korea, only data from the United States were analyzed because there was no data for each radionuclide that were disclosed when disposing of radioactive waste in LILW repository and there was no nuclear power plant that had been decommissioned. Correlation analysis of solid radioactive waste was performed by linking radioactivity of radionuclides, volume of waste, and total radioactivity data based on decommissioning work and accident data after permanent shutdown or during operation. The correlation analysis of total radioactivity, volume, and radioactivity of each nuclide of solid radioactive waste during operation and after permanent shutdown was performed using XLSTAT, an Excel add-in software, for carrying out Mann-Kendall Test and estimating Sen’s slope. Trends during operation and after permanent shutdown were compared and the effects of specific events or tasks were analyzed. This study is expected to be utilized as basic data related to safety management of decommissioning Korea nuclear power plants in future.

Waste that contains or is contaminated with radionuclides arises from a number of activities involving the use of radioactive material. Such activities include the operation and decommissioning of nuclear facilities; the use of radionuclides in medicine, industry, agriculture, research and education. Radioactive waste must be safely disposed in a radioactive waste repository for the protection of public health and the environment. In order to safely dispose of radioactive waste in a repository, it is important to derive an optimal predisposal management scenario because radioactive waste must be processed (i.e. processing (pretreatment, treatment and conditioning), storage and transport) for satisfying waste acceptance criteria (WAC). Optimal scenario of predisposal management of radioactive waste is derived for considering the balancing of exposures of workers and/or those of members of the public, the short term and long term risk implications of different waste management strategies, the technological options available and the costs. However, existing studies for deriving the optimal scenario of predisposal management of radioactive waste have evaluated only the radiation dose of workers and public within given scenarios using fixed value, or have derived optimal single process (i.e. decontamination) of predisposal management using Multi-Attribute Decision Making (MADM) methodology. In this study, optimal predisposal management scenario is derived by evaluating exposures of workers using system dynamics (SD) technique. Radiation dose assessment SD model was modeled using VENSIM® code developed by VENTANA systems Inc.. SD Model has the advantage of being able to respond flexibly when decision makers want to change input data and it has the advantage of being able to track dynamically changing phenomena and visually confirm interdependence. After that, based on the SD model derived from this study, evaluations of exposures of public, cost, and technicality will be added to be utilized when establishing an optimal scenario of predisposal management of radioactive waste considering multi attribute.