The Korea Atomic Energy Research Institute (KAERI) has facilities that are operated for the purpose of treating radioactive wastes and storing drums before sending them to a disposal site. Domestic regulations related to nuclear facility require radiological dose assessment resulting from release of gaseous radioactive effluent of nuclear facilities. In this study, ICRP-60-based dose conversion factors were applied to evaluate the radiation dose to residents in the event of operation and accident for the radioactive waste management facilities in KAERI. The radioactive gaseous effluent generated from each facility diffuse outside the exclusion area boundary (EAB), causing radiation exposure to residents. To evaluate the external exposure dose, the exposure pathways of cloudshine and radioactive contaminated soil were analyzed. The internal exposure dose was estimated by considering the exposure from respiration and ingestion of agricultural and livestock products. The maximum individual exposure dose was evaluated to be 1.71% compared to the dose limit. The assumed situation used for accidental scenarios are as follows; A fire inside the facility and falling of radioactive waste drum. It was a fire accident that caused the maximum exposure dose to individual and population living within an 80 km radius of the site. At the outer boundary of the low population zone (LPZ), the maximum effective dose and thyroid equivalent dose were estimated as 8.92 E-06% and 5.29 E-06%, respectively, compared to the dose limit. As a result of evaluating the radiological exposure dose from gaseous emissions, the radioactive waste treatment facilities and its supplementary facilities meet the regulations related to nuclear facility, and are operated safely in terms of radiological environmental impact assessment.

KAERI has developed a Radioactive Waste Information Management System (RAWINGS) to manage the life-cycle information from the generation to the disposal of radioactive waste, in compliance with the low- and medium-level radioactive waste acceptance criteria (WAC). In the radioactive waste management process, the preceding steps are to receive waste history from the waste generators. This includes an application for a specified container with a QR label, pre-inspection, and management request. Next, the succeeding steps consist of repackaging, treatment, characterization, and evaluating the suitability of disposal, for a process to transparently manage radioactive wastes. Since the system operated in 2021, The system is enhanced to manage dynamic information, including the tracking of the location of radioactive waste and the repackaging process. Small packages of waste could be classified as either radioactive or clearance waste during pre-inspection. Furthermore, waste generated in the past has already been packaged in drums, and a new algorithm has been developed to apply the repackaging when reclassification is required. All radioactive waste with the unique ID number on the specific container is managed within a database, the total amount and history of waste are managed, and statistical information is provided. This system is continuously be operated and developed to oversee life-cycle information, and serve as the foundational database for the Waste Certification Program (WCP).

As the acceptance criteria for low-intermediate-level radioactive waste cave disposal facilities of Korea Radioactive Waste Agency (KORAD) were revised, the requirements for characterization of whether radioactive waste contains hazardous substances have been strengthened. In addition, As the recent the Nuclear Safety and Security Commission Notice (Regulations on Delivery of Low- Medium-Level Radioactive Waste) scheduled to be revised, the management targets and standards for hazardous substances are scheduled to be specified and detailed. Accordingly, the Korea Atomic Energy Research Institute (KAERI) needs to prepare management methods and procedures for hazardous substances. In particular, in order to characterize the chemical requirements (explosiveness, ignitability, flammability, corrosiveness, and toxicity) contained in radioactive waste, it must be proven through documents or data that each item does not contain hazardous substances, and quality assurance for the overall process must be provided. In order to identify the characteristics of radioactive waste that will continue to be generated in the future, KAERI needs to introduce a management system for hazardous substances in radioactive waste and establish a quality assurance system. Currently, KAERI is thoroughly managing chelates (EDTA, NTA, etc.), but the detailed management procedures for hazardous substances related to chemical requirements in radioactive waste in the radiation management area specified above are insufficient. The KAERI’s Laboratory Safety Information Network has a total periodic regulatory review system in place for the purchase, movement, and disposal of chemical substances for each facility. However, there is no documents or data to prove that the hazardous substances held in the facility are not included in the radioactive waste, and there are no procedures for managing hazardous substances. Therefore, it is necessary to establish procedures for the management of hazardous substances, and we plan to prepare management procedures for hazardous substances so that chemical substances can be managed according to the procedures at each facility during preliminary inspection before receiving radioactive waste. The procedure provides definitions of terms and types of management targets for each characteristic of the chemical requirements specified above (explosiveness, ignition, flammability, corrosiveness, and toxicity). In addition, procedure also contains treatment methods of radioactive waste generated by using hazardous substances and management methods of in/out, quantity, history of that substances, etc. As the law is revised in the future, management will be carried out according to the relevant procedures. In this study, we aim to present the hazardous substance management procedures being established to determine whether radioactive waste contains hazardous substances in accordance with the revised the notice and strengthened acceptance criteria. Through this, we hope to contribute to improving reliability so that radioactive waste could be disposed of thoroughly and safely.

Every engineering decision in radioactive waste management should be based on both technical and economic considerations. Especially, the management of low-level radioactive waste (LLW) is more critical on economic concerns, due to its long-term and continuous nature, which emphasizes the importance of economic analysis. In this study, economic factors for LLW management were discussed with appropriate engineering applications. Two major factors that should be taken into account when assessing economic expectations are the accuracy of the results and its proper balancing with ALARA philosophy (As Low As Reasonably Achievable). The accuracy of the results depends on the correct application of alternatives within a realistic framework of waste processing. This is because the LLW management process involves variables such as component type, physical dimensions, and the monetary value at the processing date. Two commonly used alternatives are the simplified lump sum present worth and levelized annual cost methods, which are based on annual and capital costs. However, these discussions on alternatives not only pertain to the time series value of operational costs but also to future technical advancements, which are crucial for engineers. As new research results on LLW treatment emerge, proper consideration and adoption should be given to technical cost management. As safety is the core value of the entire nuclear industry, the ALARA philosophy should also be considered in the cost management of LLW. The typical cost of exposure in man-rem has ranged from $1,000 to $20,000 over the past decades. However, with increasing concerns about health and international political threats, the cost of man-rem should be subject to stricter criteria, even the balancing of costs and safety concerns is much controverse issue. Throughout the study, the importance of incorporating proper engineering insights into the assessment of technical value for the financial management of LLW was discussed. However, it’s essential to remember that financial management should not be solely assessed based on the size of expenses but rather by evaluating the current financial status, the value of money at the time, and anticipated future costs, considering the specific context and timeframe.

The radionuclide management process is a conditioning technology to reduce the burden of spent fuel management, and refers to a process that can separate and recover radionuclides having similar properties from spent fuels. In particular, through the radionuclide management process, high heat- emitting, high mobility, and high toxicity radionuclides, which have a significant impact on the performance of disposal system, are separated and managed. The performance of disposal system is closely related to properties (decay heat and radioactivity) of radioactive wastes from the radionuclide management process, and the properties are directly linked to the radionuclide separation ratio that determines the composition of radionuclides in waste flow. The Korea Atomic Energy Research Institute have derived process flow diagrams for six candidates for the radionuclide management process, weighing on feasibility among various process options that can be considered. In addition, the GoldSim model has been established to calculate the mass and properties of waste from each unit process of the radionuclides management process and to observe their time variations. In this study, the candidates for the radionuclide management process are evaluated based on the waste mass and properties by using the GoldSim model, and sensitivity analysis changing the separation ratio are performed. And the effect of changes in the separation ratio for highly sensitive radionuclides on waste management strategy is analyzed. In particular, the separation ratio for high heat-emitting radionuclides determines the period of long-term decay storage.

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. RadiagemTM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 μSv·h−1. The range of the mean annual average personnel dose equivalent is 0.41–2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

Natural radionuclides-containing substances (NORM) contain natural radionuclides and cause radiation exposure. In Korea, safety management measures were needed to deal with and dispose of radon mattresses containing monazite in relation to such NORM. However, there is no clear safety management system related to NORM waste in Korea. In order to manage this reasonably and systematically, it is necessary to investigate and analyze standards and management measures related to the treatment and disposal of NORM waste. Therefore, this study investigated and analyzed the exemption and clearance level of NORM waste regulations in international organizations and foreign countries. IAEA GSR Part 3, 2013/59/Euratom, ANSI/HPS N13.53, CRCPD SSRCR Part N, and ARPANSA Publications 15 safety management regulations were analyzed to investigate safety management standards for NORM waste. The exemption and clearance level in international organizations and foreign countries were compared and analyzed based on radioactive concentration and dose. In addition, the management measures proposed for each literature were also investigated. As a result of the analysis, IAEA GSR Part 3 applied 1 mSv as a regulatory exemption level, 1 Bq/g for uranium and thorium series as a clearance level, and 10 Bq/g for K-40 nuclides. The IAEA recommends a differential approach to the potential and scale of exposure. The EU applied 1 Bq/g to uranium and thorium families and 10 Bq/g to K-40 nuclides for both regulatory exemption and clearance levels. The EU recommended that it be managed in proportion to the scale and likelihood of exposure as a result of the action. It is analyzed that this is similar to the IAEA’s management plan. In the United States, there was no single federal government radioactive concentration and dose for NORM management. The management plan differed in management status and level from state to state, and K-40 was excluded from regulation unless it was intentionally enriched. In the case of Australia, the radioactive concentration of uranium and thorium was 1 Bq/g as a standard for regulatory exemption and 1 mSv as a dose. As a management plan, it was suggested to dispose of waste by means of accumulation, dilution/dispersion, and reclamation. It was also suggested that the scale of exposure, like international organizations, take into account the possibility. The results of this study are believed to be used as basic data for presenting domestic NORM waste treatment and disposal methods in the future.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

The amount of waste that contains or is contaminated with radionuclides is increasing gradually due to the use of radioactive material in various fields including the operation and decommissioning of nuclear facilities. Such radioactive waste should be safely managed until its disposal to protect public health and the environment. Predisposal management of radioactive waste covers all the steps in the management of radioactive waste from its generation up to disposal, including processing (pretreatment, treatment, and conditioning), storage, and transport. There could be a lot of strategies for the predisposal management of radioactive waste. In order to comply with safety requirements including Waste Acceptance Criteria (WAC) at the radioactive waste repository however, the optimal scenario must be derived. The type and form of waste, the radiation dose of workers and the public, the technical options, and the costs would be taken into account to determine the optimal one. The time required for each process affects the radiation dose and respective cost as well as those for the following procedures. In particular, the time of storing radioactive waste would have the highest impact because of the longest period which decreases the concentrations of radionuclides but increases the cost. There have been little studies reported on optimization reflecting variations of radiation dose and cost in predisposal management scenarios for radioactive waste. In this study, the optimal storage time of radioactive waste was estimated for several scenarios. In terms of the radiation dose, the cumulative collective dose was used as the parameter for each process. The cost was calculated considering the inflation rate and interest rate. Since the radiation dose and the cost should be interconvertible for optimization, the collective dose was converted into monetary value using the value so-called “alpha value” or “monetary value of Person-Sv”.

During the operation of the nuclear power plant, various radioactive waste are generated. The spent resin, boron concentrates, and DAW are classified as a generic radioactive waste. They are treated and stored at radioactive waste building. In the reactor vessel, different types of radioactive waste are generated. Since the materials used in reactor core region exposed to high concentration of neutrons, they exhibit higher level of surface dose rate and specific activity. And they are usually stored in spent fuel pool with spent fuel. Various non-fuel radioactive wastes are stored in spent fuel pool, which are skeleton, control rod assembly, burnable neutron absorber, neutron source, in core detector, etc. The skeleton is composed of stainless 304 and Inconel-718. There are two types of control rod assembly, that are WH type and OPR type. The WH type control rod is composed of Ag-In-Cd composites. The OPR type control rod is composed of B4C and Inconel-625. In this paper, the characteristics and storage status of the non-fuel radioactive waste will be reported. Also, the management strategy for the various non-fuel radioactive waste will be discussed.

After the Fukushima accident, significant amount of radioactively contaminated waste has been generated with 50~250 m3/day and stored in tanks of the Fukushima Daiichi nuclear power plant site. The contaminated water is treated by various treatment facility such as KURION, SARRY, Reverse Osmosis, and ALPS to remove 62 radioactive nuclides except H-3. For the contaminated water treatment process, massive secondary wastes such as sludge, spent adsorbent, and so on as by-product are being generated by the facilities. In Japan, to treat the secondary wastes, melting technologies such as GeoMelt, In-can vitrification and Cold Crucible Induction Melting vitrification are considered as a candidate technologies. In this study, the technologies were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for treatment of the secondary wastes.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

Korea currently has two permanent shutdown Nuclear Power Plants (NPPs), and the decommissioning project is expected to begin soon, starting with the first commercial NPP. The decommissioning project will eventually be the disposal of radioactive waste in the final stage of the work, and in that respect, proper tracking and history management should be well established in the management of waste. This is in line with the guidelines that regulatory agencies should also properly manage radioactive waste. Therefore, this study intends to examine the factors that should be considered in terms of tracking and management of radioactive waste in decommissioning nuclear facilities. The starting and final point of tracking radioactive waste generated during decommissioning is the physical inventory of the current as-is state and the final container. In this respect, the tracking of waste starts from the beginning of the dismantling operation. Thus, at the stage of approval of the decommissioning work, it may begin with an ID scheme, such as the functional location in operation for the target System, Structure, and Components (SSCs). As the dismantling work progresses, SSCs will be classified by nature and radiological level, which will be placed in containers in small packaging units. At this time, the small package should be given an ID. After that, the dismantling work leads to the treatment of waste, which involves a series of operations such as cutting, decomposition, melting, and decontamination. Each step in which these tasks are performed will be placed in a container, and ID assignment is also required. Until now, the small packaging container is for transfer after each treatment, and it is placed in the storage container in the final stage, at which time the storage container also gives a unique ID. Considerations for follow-up management were reviewed assuming solid waste, which is the majority of dismantled radioactive waste considered in this study. The ID system should be prepared from the start of the dismantling work, ID generation of the small transporting container and ID generation of the final disposal container during the intermediate waste treatment process, and each ID generation of the previous stage should be linked to each generation stage. In addition, each ID must be generated, and the definition of the grant scheme and attributes is required.

Korea Atomic Energy Research Institute is developing a radionuclide management processes as a conditioning technology to reduce the burden of spent fuel disposal. The radionuclide management process refers to a process managing radionuclides with similar properties by introducing various technology options that can separate and recover radionuclides from spent fuels. In particular, it is a process aimed at increasing disposal efficiency by managing high-heat, high-mobility, and high-toxic radionuclides that can greatly affect the performance of the disposal system. Since the radionuclide management process seeks to consider various technology options for each unit process, it may have several process flows rather than have a single process flow. Describing the various process flows as a single flow network model is called the superstructure model. In this study, we intend to develop a superstructure model for the radionuclide management process and use it as a model to select the optimal process flow. To find the optimal process flow, an objective function must be defined, and at the fuel cycle system level multiple objectives such as effectiveness (disposal area), safety (explosure dose), and economics (cost) can be considered. Before performing the system-level optimization, it is necessary to select candidates of process flow in consideration of waste properties and process efficiency at the process level. In this study, a sensitivity analysis is conducted to analyze changes in waste properties such as decay heat and radioactivity when the separation ratio varies due to the performance change for each unit process of the radionuclide management process. Through this analysis, it is possible to derive a performance range that can have waste properties suitable for following waste treatment, especially waste form manufacturing. It is also possible to analyze the effect of waste properties that vary according to the performance change on waste storage and management approaches.

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

The decommissioning of nuclear power plant (NPP) consists of various activities, such system decontamination, take out of activated components, segmentation of the activated components, site remediation, etc. During various activities, the generation of radioactive wastes and radiation exposure to workers is expected. The systematic waste management during the activities is important to implement the decommissioning. The inefficient waste management usually bring significant delay in decommissioning process and results in increase of decommissioning cost. The radiation exposure management is also an important issue. It is generally accepted that the hot spot, generated from operation and decommissioning of NPP, is observed in many places within containment building. Although the health physicists measure the radiation in various points, the unintended hot spots are sometimes generated and observed. The effective radiation exposure management also requires the control of personnel and space during various activities. In this study, the radiation exposure and waste management experiences of Zion NPP is reviewed. The primary nuclides and radiation exposure during various activities are systematically studied to achieve the main objectives of this paper.

Solid radioactive waste such as rubble, trimmed trees, contaminated soil, metal, concrete, used protective clothing, secondary waste, etc. are being generated due to the Fukushima nuclear power plant accident occurred on March 11, 2011. Solid radioactive waste inside of Fukushima NPP is estimated to be about 790,000 m3. The solid radioactive waste includes combustible rubble, trimmed trees, and used protective clothing, and is about 290,000 m3. These will be incinerated, reduced to about 20,000 m3 and stored in solid waste storage. The radioactive waste incinerator was completed in 2021. About 60,000 m3 of rubble containing metal and concrete with a surface dose rate of 1 mSv/h or higher will be stored without reduction treatment. Metal with a surface dose rate of 1 mSv/h or less are molten, and concrete undergoes a crushing process. About 60,000 m3 of contaminated soil (0.005 ~1 mSv/h) will be managed in solid waste storage without reduction treatment. The amount of secondary waste generated during the treatment of contaminated water is about 6,500 huge tanks, and additional research is being conducted on future treatment methods.

To efficiently manage the waste packages like drums, it is meaningful to utilize the data of placement and emplacement of disposed waste in geological storage. For the transparent and real-time management of disposal data, both technical as well as administrative factors must be included. To this end, MIRAE-EN Co., Ltd. has developed a radioactive waste tracking and management system (m-trekⓇ v1.0) through radioactive waste management life cycle which is supported by KETEP. Enhancing the functional features of m-trekⓇ, IoT-based drum location measuring and data of those drums, such as position, radionuclides, activity, and dose etc., are to be collected and monitored through data modeling and visualization, which might be utilized in emplacing the loaded drums according to specifically certain criteria of internal and external data of disposal site. Position measuring using Beacon utilizes Received Signal Strength Indicator (RSSI) to locate the correct position in 3D area. Since RSSI is affected by the surrounding environment, it is required to corrective optimization. In addition, error and deviation of previously applied Gaussian filter method, was corrected and improved through AI learning model. According to this location information and those data, the prototype in future provides the visualization of drum position and its relevant data for administrative purpose such as monitoring, emplacement and other management policy.

The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.