Wolsong unit 1 (W1), which is a CANDU-6 type PHWRs that had been operated for 30 years since 1983, was shutdown in 2019. In this study, the radioactive waste levels of calandria and concrete structures were calculated to establish a decommissioning plan for W1. The specific systems within the scope of this study were grouped into 6 major categories as follows: Calandria, End Shield, Fuel Channel Assembly, Reactivity Control Device, End Shield Support, Vault. The main operating history of W1 is that the re-tubing project was performed. These characteristics and operation history were reflected in the evaluation. The neutron flux and energy spectrum of each structure were calculated by using MCNP code, and ORIGEN code is implemented to the calculation of radioactivity for each nuclide using the results from MCNP and the material information of the structure. As for the impurity information, ASTM B350, B351, B353 standard was used for zircaloy alloy. For other alloy, impurity information provided by NUREG/CR-3474 was applied. Since W1 is expected to be decommissioned immediately, the waste level was evaluated under cooling conditions for 5 years after permanent shutdown. Through the level evaluation of each component obtained as a result of the study, it can be used as basic data for the radioactive waste management of the decommissioning plan.

The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

Source localization technique using acoustic emission (AE) has been widely used to track the accurate location of the damaged structure. The principle of localization is based on signal velocity and the time difference of arrival (TDOF) obtained from different signals for the specific source. However, signal velocity changes depending on the frequency domain of signals. In addition, the TDOF is dependent on the signal threshold which affects the prediction accuracy. In this study, a convolutional neural network (CNN)-based approach is used to overcome the existing problem. The concrete block corresponding to 1.3×1.3×1.3 m size is prepared according to the mixing ratio of Wolseong low-to-intermediate level radioactive waste disposal concrete materials. The source is excited using an impact hammer, and signals were acquired through eight AE sensors attached to the concrete block and a multi-channel AE measurement system. The different signals for a specific source are time-synchronized to obtain TDOF information and are transformed into a time-frequency domain using continuous wavelet transform (CWT) for consideration of various frequencies. The developed CNN model is compared with the conventional TDOF-based method using the testing dataset. The result suggests that the CNN-based method can contribute to the improvement of localization performance.

In order to indirectly evaluate the inventory of difficult-to-measure (DTM) nuclides in radioactive waste, the scaling factor method by key nuclide has been used. It has been usually applied to low-and intermediate-level dry active waste (DAW), and the tolerance of 1,000% margin of error in the US, that is the factor of 10, is applied as an allowable confidence limits considering the inhomogeneity of the waste and the limitation of sample size. This is because the scaling factor method is based on economic efficiency. Confidence limits is the uncertainty (sampling error) according to predicting the mean value of the population by the mean value of the sample at 95% confidence level, reflecting the limitations of sample size (representation) with the standard deviation. If the standard deviation is large, the sample size can be increased to satisfy the allowable confidence limits. In the new nuclear power plants, the concentration of cesium nuclide (137Cs) in radioactive waste tends to be very low due to advances in nuclear fuel and reactor core management technology, which makes it very difficult to apply cesium as a key nuclide. In addition, it is inevitable to apply the mean activity concentration method, which reasonably and empirically derives the concentration of DTM nuclides regardless of key nuclide, when the correlation between key and DTM nuclides is not significant. The mean activity method is a methodology that applies the average concentration of a sample set to the entire population, and is similar to applying the average concentration ratio between key and DTM nuclides of a sample set to the population in the scaling factor method. Therefore, in this paper, the maximum acceptable uncertainty (confidence limits) at a reasonable level was studied when applying the mean activity concentration method by arithmetic mean unlike the scaling factor method which usually uses the geometric mean method. Several measures were proposed by applying mutatis mutandis the acceptable standard deviation in radiation measurement and the factor of 10 principle, etc., and the appropriateness was reviewed through case analysis.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

Low to intermediate radioactive waste disposal concrete structures are subjected to coupled hydromechanical conditions and the identification of structural damage is crucial to ensure safe long-term disposal. Different damage models for concrete and the surrounding rock can affect the damage characteristics of radioactive waste disposal structures. In this study, the effects of different rock damage models are applied to the hydro-mechanical-damage coupled structural analysis of the Wolseong Low and Intermediate Level Radioactive Waste Disposal Center silo. A two-dimensional model of the disposal silo was modeled using the finite element analysis software COMSOL and the Mazars’ damage model was applied to the silo concrete. The Mazars’ model parameters were obtained from uniaxial compression and tensile tests on cylindrical concrete specimens after 28 days of water curing and further 32 days of wet curing at 75°C). The COMSOL embedded Richards equation module was used to simulate hydraulic analysis. Structural loading due to waste disposal was applied at the bottom of the silo structure and the damage evolution characteristics were investigated. The non-linear mechanical rock behavior obtained from laboratory tests (Hoek-Brown criterion, resonant column test, Mazar’s damage model) and field tests (Goodman Jack) were input to assess the effects of different rock damage models. The results highlight the importance of structural damage consideration when assessing the long-term stability and safety of underground radioactive waste disposal structures under coupled hydro-mechanical conditions.

Near-surface disposal facility is more susceptible to intrusion than underground repository, resulting in more possible pathways for contaminant release. Alike human intrusion, animals (e.g. Ants, Moles, etc.) could intrude into the disposal site to excavate burrows, which could cause direct release of contaminants to biosphere. In this paper, animal intrusion is demonstrated using GoldSim’s commercial contaminant transport module and impact on the integrity of the near-surface disposal facility is evaluated in terms of fractional release rate of the contaminants. In this study, the near-surface disposal facility is modelled with a single concrete vault to contain radionuclide according to LLW concentration limit stated in NSSC notice No.2020-6. The release of contaminants is modelled to occur directly after the institutional control period, and the contaminants are mostly transported from the concrete vault to cover layers via diffusion. To produce mathematical model of the release of the contaminants due to animal intrusion, firstly, the fraction of burrow volume for each cover layer is calculated separately for each animal species, based on their maximum possible intrusion depth. In this study, fractions of burrow volume for ants and moles are calculated based on their maximum possible intrusion depths, where for ants is 2–3 m, and for moles is 0.1–0.135 m. Then, assuming that the contaminants are distributed homogeneously throughout each cover layers by diffusion, fraction of contaminants transported into the uppermost layer via excavation of the burrow is calculated for each layer based on burrow volume, and fraction of contaminants removed from the uppermost layer to the layers below via collapse of the burrow is also calculated based on the burrow volume. Lastly, the net transportation of contaminants into and out of the burrow via excavation and collapse, respectively, is calculated and demonstrated using direct transfer rate function of the GoldSim. Based on the simulated result, the maximum mass flux is too minor to cause a meaningful impact on the safety. The peak mass flux of the most sensitive radionuclide, I-129, is witnessed at around year 1,470, with a flux value of 5.36×10−6 g·yr−1. This minor release of the contaminants could be due to cover layers being much thicker than the maximum possible intrusion depth of the animals, preventing the animal intrusion into the deeper layers of higher radionuclide concentration. In future, this study can be used to provide a guidance and fundamental data for scenario development and safety evaluation of the near-surface disposal facility.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

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.

Low-and intermediate level waste (LILW) should be solidified and satisfy the waste acceptance criteria (WAC) to be disposed of in the LILW repository. The LILW should be uniformly solidified and should maintain its structural stability under the expected condition according to the WAC. Compressive strength of cement solidified waste should satisfy at least 3.44 MPa to be disposed of in the repository. In addition, its compressive strength should satisfy at least 3.44 MPa after the irradiation, immersion and leaching test. The compressive strength test and dimension of test specimen differ according to countries. However, measured compressive strength of solidified waste is affected by geometry of specimen and test condition. Diameter, ratio between diameter and height, and porosity are one of factors that affect to the compressive strength of cement solidified waste. Generally, specimen with larger diameter shows higher value of measured compressive strength. The ratio of height and diameter shows similar tendency to the diameter while larger porosity generally lowers the compressive strength. In other hands, higher compressive strength is expected when the loading rate is higher during the compressive strength test. U.S. is applying loading rate from ASTM C39 (0.25±0.05 MPa) for the compressive strength test while Korea is applying loading rate from KS F 2405 (0.6 MPa·s−1). France applies loading rate following FT-02-010 (0.5 MPa·s−1) for cement solidified waste. As the measured compressive strength increases when the loading rate increases, the effect of loading rate to the compressive strength of cement solidified waste should be assessed by quantification and consider its effect on the sight of regulation. In this study, the effect of geometric parameters of specimen and test condition to the compressive strength are checked by manufacturing specimen by solidifying mock sludge waste with cement. To prevent increasing amount of secondary waste, effects of ratio of height and diameter and porosity to the compressive strength are checked while diameter value is fixed. For loading rate, loading rate from ASTM C39 and KS F 2405 were compared. Existence of significant variance of measured compressive strengths of cement solidified waste are check by performing statistical analysis. Finally, by analyzing the relationship between test condition and measured compressive strength, the test method that measures the compressive strength conservatively is aimed to be derived.

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.

In Korea, it is expected that the decommissioning of nuclear reactors will increase due to the license termination of reactors constructed in the 1960s to the 80s. According to the investigation of KORAD, VLLW accounts for 67.10% of decommissioning wastes and amounts to about 413,336 drums. Due to their huge amount, it is necessary to create an appropriate decommissioning waste management plan even though VLLW is disposed at the second-phase disposal facility of the Gyeongju repository. For efficient reduction in decommissioning wastes, it is required to actively use a clearance of metallic and concrete radioactive wastes. Regulations of nuclear safety and security commission notice that the radioactive waste can be reused or recycled if it meets the clearance criterion, 10 μSv·y−1 for individual dose. Therefore, it is important to develop a computational code which calculate individual doses for each scenario, and determine whether the clearance criterion is satisfied. However, in the case of metallic waste, RESRAD-RECYCLE used in dose assessment for the clearance has no longer been maintained or updated since 2005 and there is no code for recycling of concrete waste. For this reason, a dose assessment code RUCAS (Recycle-Underlying Computational dose Assessment System) has been developed by Ulsan National Institute of Science and Technology (UNIST). A point kernel method is adopted into external dose assessment model to calculate more realistic options, which are various geometries of source, and shielding effect. In the case of internal radiation, equations of internal dose from IAEA are used. This research conducts a verification of dose assessment model for recycling of metallic radioactive waste. RESRAD-RECYCLE is the comparison object and results from RESRAD-RECYCLE validation report are referenced. Targets are 14 recycling scenarios composed up to the smelting metal step of four steps, which are arising scrap metal, smelting scrap metal, and fabrication of metal product, and reusing/recycling of product. Seven isotopes, which are Ac-227, Am-241, Co-60, Cs-137, Pu-239, Sr- 90, and Zn-65, are selected for calculation. Validation results for external dose vary by isotopes, but show acceptable differences. It seems to be caused by difference in the calculation method. In the case of internal dose using same calculation formula, results are exactly matched to RESRAD-RECYCLE for all isotopes. Consequently, RUCAS can conduct functions supported by RESRAD-RECYCLE well and future work will be conducted related to domestic recycling scenarios considering public acceptance, and verification with radiation shielding codes for various geometries of source.

Glass fiber, which was used as an insulation material in pipes near the steam generator system of nuclear power plants, is brittle and the size of crushed particles is small, so glass fiber radioactive waste (GFRW) can cause exposure of workers through skin and breathing during transport and handling accidents. In this study, Q-system which developed IAEA (International Atomic Energy Agency) for setting the limit of radioactivity in the package is used to confirm the risk of exposure due to an accident when transporting and handling GFRW. Also, the evaluated exposure dose was compared with the domestic legal effective dose limit to confirm safety. Q-system is an evaluation method that can derive doses according to exposure pathway (EP) and radioactivity. Exposure doses are calculated by dividing into five EP: QA, QB, QC, QD, and QE. Since the Q-system is used to set the limit of radioactivity that the dose limits is satisfied to nearby workers even in package handling accidents, the following conservative assumptions were applied to each EP. QA, QB are external EP of assuming complete loss of package shielding by accident and radiation are received for 30 minutes at 1 m, QC is an internal EP that considers the fraction of nuclides released into the air and breathing rate during accident, and QD is an external EP that skin contamination for 5 hours. Finally, QE is an internal and external EP by inert gases (He, Ne, Ar, Kr, Xe, Rn) among the released gaseous nuclides, but the QE pathway was excluded from the evaluation because the corresponding nuclide was not present in the GFRW products used for evaluation. In this study, the safety evaluation of GFRW was performed package shielding loss and radioactive material leakage due to single package accident according to assumption of four pathways, and the nuclide information used the average radioactivity for each nuclide of GFRW. As a result of the dose evaluation, QA was evaluated as 2.73×10−5 mSv, QB as 1.06×10−6 mSv, QC as 7.53×10−3 mSv, and QD as 2.10×10−6 mSv, respectively, and the total exposure dose was only 7.56×10−3 mSv, it was confirmed that when compared to the legal limits of the general public (1 mSv) and workers (20 mSv) 0.756% and 0.038%, respectively. In this study, it was confirmed that the legal limitations of the general public and workers were satisfied evens in the event of an accident as a result of evaluating the exposure dose of nearby targets for package shielding loss and radioactive material leakage while transporting GFRW. In the future, the types of accidents will be subdivided into falling, fire, and transportation, and detailed evaluation will be conducted by applying the resulting accident assumptions to the EP.

It has been discovered that the isosaccharinic acid (ISA) formed in a cellulose degradation leachate were capable of forming soluble complexes with thorium, uranium (IV) and plutonium. Since 1993, the ISA has received particular attention in the literature due to its ability to complex a range of radionuclides, potentially affecting the migration of radionuclides. ISA is formed as a result of interactions between cellulosic materials within the waste inventory and the alkalinity resulting from the use of cementitious materials in the construction of the repository. In an alkaline cementitious environment, cellulose degrades mainly via a peeling-off reaction. The main degradation product is ISA, a polyhydroxy type of ligand forming stable complexes with tri- and tetravalent radionuclides. ISA can have an adverse effect on the sorption of radionuclides to an extent which depends on its concentration in the cement pore water and potentially enhance their mobility. The concentration of ISA is governed by several factors such as cellulose loading, cement porosity, extent of cellulose degradation, etc. The sorption of ISA on cement, however, is the process which governs the concentration of ISA in the pore water. According to the experimental result from a literature, the ISA concentration in facilities with a cellulose loading of 5% is calculated to be of the order of 10−4 M. At this level, the effect of cellulose degradation products on radionuclide sorption is negligibly small. Recently in Korea, cellulous limits as waste acceptance criteria is studying and planning to prepare the detailed requirement for near surface radioactive waste disposal facilities. It is desirable to suggest consideration on cellulose disposal limits around the time that the regulatory body and concern organizations establish the cellulose disposal limits as follows. Firstly, identify the cellulose effect on the sorption of the nuclides as cementitious disposal environments such as affected nuclides, threshold value and contribution to radiological risks under domestic disposal environment. Secondly, make sure and consider the difference between lab-scale experimental conditions and probability occurring in real disposal conditions such as probability for generation and persistence of pH in cellulosic material disposal conditions and cellulosic material disposal methods. Finally, consider characterization of cellulosic material such as polymerization, contents of cellulose in law material and time of degradation process. As a result, desirable cellulose limits are to set up for both safety and economic aspect.

Immobilization of radioactive borate waste containing a high boron concentration using cement waste form has been challenged because the soluble borate phase such as boric acid reacts with calcium compounds, hindering the hydration reaction in cement waste form. Metakaolin-based geopolymer waste form which has a pure aluminosilicate system without calcium can be a promising alternative for the cement; however, secondary B-O-Si networks are formed by a reaction between borate and silicate, resulting in poor mechanical characteristics such as low compressive strength and final setting retardation. Thus, it is important to optimize the Si/Al molar ratio and curing temperature which are critical parameters of geopolymer waste form to increase borate waste loading and enhance the durability of geopolymer. Here, metakaolin-based geopolymer waste form to immobilize simulant radioactive borate waste was fabricated by varying the Si/Al molar ratio and curing temperature. The 7 days-compressive strength results reveals that the Si/Al molar ratio of 1.4 and curing at 60°C is advantageous to achieving high waste loading (30wt%). In addition, geopolymer waste forms with the highest borate waste loading exceeded the 3.445 MPa after the waste form acceptance criteria such as thermal cycling, gamma irradiation, and water immersion tests. The leachability index of boron was 7.56 and the controlling leaching mechanism was diffusion. The thermal cycling and gamma irradiation did not significantly change the geopolymer structure. The physically incorporated borate waste was leached out from geopolymer waste form during leaching and water immersion tests. Considering these results, metakaolin-based geopolymer waste form with a low Si/Al ratio is a promising candidate for borate waste immobilization, which has been difficult using cement.

With the development of the nuclear industry and the increase in the use of radioactive materials, the generation of radioactive waste is increasing. As the generation of radioactive waste increases, the occurrence of related safety accidents is also increasing, and it is necessary to develop a radioactive waste monitoring technology to prevent such accidents in advance and efficiently manage radioactive waste. In Information and Communication Technology (ICT), various ICT technologies such as Internet of Things (IoT), Augmented Reality (AR), and Virtual Reality (VR) that can help with the safety management of these radioactive wastes are being developed. In this study, a radioactive waste monitoring technology was developed using ICT technology, such as management of the entire cycle history of waste using Quick Response (QR) codes, and development of AR visualization technology for small packages of radioactive waste. In addition, by using IoT technology to collect desired data from sensors and store the results, after the waste drum is loaded in the waste storage, a technology was developed to track and monitor the history and movement of the waste drum from repackaging to transfer to the storage. The data required for monitoring the radioactive waste drum includes location information, whether the drum is open or closed, temperature and humidity, etc. To collect this information, a drum monitoring technology was built with a 2.4 G wireless router, an anchor constituting a virtual zone, a tag to be mounted on the drum container, and a WNT server that collects sensor data. The network tool provided by WirePas was used for network configuration, and the status of gateways and nodes can be monitored by interworking with the WNT server. The configured IoT sensor technology were tested in a waste storage environment. Four anchors were installed and linked to the network to match the virtual zone and the real storage zone, and it was confirmed whether the movement of the tag was recorded on the network while moving the tag including the IoT sensor for analyzing location information. Based on these research results, it can contribute to the safety management of radioactive waste and establishment of Waste Acceptance Criteria (WCP) by and managing the history and monitoring the waste in the entire cycle from repackaging to disposal.

During decommissioning of a nuclear power plant, a large amount of radioactive waste is produced, and it is known to cost more than 300 billion won to dispose the waste. To reduce the disposal cost, it is essential to minimize the number of radioactive waste drums, which can be achieved by detecting and removing hotspot contaminations in the radioactive waste drums. Therefore, a Compton CT system for radioactive waste monitoring is under development, which provides the images of both the internal structure of the drum and the radioactive hotspot(s) in the drum. Based on the acquired information, the activity of hotspots can be estimated. The performance of the system is affected by various geometry factors. Therefore, it is essential to determine optimal configuration by evaluating the effects of the factors on the performance of the system. In the present study, we determined the optimum value of the factors and then predicted the performance of the optimized system by using a simulator based on the Geant4 Monte Carlo simulation. For optimization, the factors were evaluated in terms of structural similarity index measure (SSIM) and measurement time. The considered factors were the activity of the CT source, source to object distance (SOD), object to detector distance (ODD), and projection angle. The simulation result showed that the activities of the CT sources were determined as 23 mCi for 137Cs and 9.6 mCi for 60Co. The optimal SOD and ODD were 180 cm and 40 cm, respectively. The optimal projection angle was evaluated as 4° since it achieves the SSIM of 0.95 faster than other projection angles. With the optimized parameters, the performance of the system was evaluated using the IAEA gamma CT standard phantom containing a hotspot of 137Cs (7.02 μCi). The Compton image was reconstructed using the back-projection algorithm, and the CT image was reconstructed using the filtered back-projection algorithm. The result showed that the location of the hotspot in the Compton image was well identified at the true position. The acquired CT image also well represented the internal structure of the phantom, and the estimated mean linear attenuation coefficient value (μ= 0.0789 cm−1) of the phantom was close to the true value (μ= 0.0752 cm−1). In addition, the hotspot activity estimated by combining the information of the Compton image and CT image was 8.06 μCi. Hence, it was found that the Compton CT system provides essential information for radioactive waste drums.

The natural barrier, a component of the deep disposal system, has site-specific characteristics depending on the site of the repository, and is one of the main considerations for long-term safety evaluation after closure along with the engineered barrier among the multiple barrier systems of the repository. The natural barrier is defined in Korea as the natural underground and surface structures that can restrict the exposure of radioactive waste, human intrusion or groundwater infiltration into a disposal facility, and the transfer of radionuclides. It includes bedrocks and soils surrounding the engineered barriers of radioactive wastes [Notice of the NSSC, No. 2020021]. This study analyzed foreign regulatory requirements related to natural barriers, requirements for natural barrier and performance target of Sweden and Finland (safety functions and target characteristics of natural barriers, e.g. natural barrier composition, geological characteristics, hydrogeological characteristics). Overseas regulations and cases referenced to derive regulations of general safety requirements on natural barrier are IAEA SSG-14, SSMFS 2008:21 in Sweden, STUK/Y/4/2018 in Finland, and POSIVA SKB Report 01, a joint report between POSIVA and SKB. The repository site and repository depth should be chosen so that the geological formation provides adequately stable and favorable conditions to ensure that the repository barriers perform as intended over a sufficient period of time. The conditions intended primarily concern temperature- related, hydrological, mechanical (for example, rock mechanics and seismology) and chemical (geochemistry, including groundwater chemistry) factors. Furthermore, the repository site should be located at a secure distance from natural resources exploited today or which may be exploited in the future [SSMFS 2008:21]. Finland regulations also suggests similar requirements [STUK Y-4-2018]. According to the above regulations, POSIVA SKB report 01 mentions both the host rock and the underground opening as natural barriers and requires a safety function, and the main safety functions of the host rock and underground opening are as follows: (1) Isolation from the surface environment; (2) Favorable thermal conditions; (3) Mechanically stable conditions; (4) Chemically favorable conditions; and (5) Favorable hydrogeological conditions with limited transport of solutes. Such safety functions would provide insight for understanding of the natural barrier of deep geological disposal system.

To decrease area of the repository for high-level radioactive waste, enhancing the disposal efficiency is needed for public acceptance. Previous studies regarding the performance assessment of KRS and KRS+ repository did not consider area-based variations of the geothermal gradient and rock thermal properties in Korea. This research estimated deposition hole spacing based on performance assessment of a repository using the distribution of geothermal gradient and rock thermal properties in Korea to increase disposal efficiency. Distributions of geothermal gradient, rock thermal properties were investigated based on 2019 Korea geothermal atlas published by Korea Institute of Geoscience and Mineral Resources (KIGAM). Effect of thermal performance parameters was analyzed using coupled thermal-hydraulic numerical simulations, and effect of rock thermal conductivity and deposition hole spacing on the maximum temperature of buffer was relatively large. In addition, distribution maps of thermal performance of a repository and deposition hole spacing were plotted using thermal performance parameters-maximum temperature of buffer regression equations and GIS data given by KIGAM. In the regions showing the highest maximum temperature of buffer in Korea, required deposition hole spacings were 10.5 m, 10.0 m, 10.1 m, respectively for KJ-II, MX-80, and FEBEX bentonite cases, and thereby additional disposal area of 40%, 33.3%, and 34.7% were required compared to that of the KRS+ repository. On the other hand, high disposal efficiency can be obtained in the regions showing the low maximum temperature of bentonite buffer. The methodology provided in this research can be used as one of the references for the selection of domestic candidate repository sites. Additional mechanical performance analysis should be conducted using distributions of mechanical properties of rock mass in Korea.

For the peaceful use of nuclear energy, the international community has devoted itself to fulfilling its obligations under the Safeguards Agreement with IAEA. In this regard, uranium in a radioactive waste drum should be analyzed and reported in terms of mass and 235U enrichment. In order to characterize radioactive wastes, gamma spectroscopy techniques can be effectively applied. In the case of high-resolution gamma spectroscopy, because an HPGe detector can provide excellent energy resolution, it can be applied to analyze a mixture having a complicated isotopic composition. However, other substances such as wood, concrete, and ash are mixed in radioactive waste with various form factors; hence, the efficiency calibration is difficult. On the other hand, In Situ Object Counting System (ISOCS) has a capability of efficiency calibration without standard materials, making it possible to analyze complex radioactive wastes. In this study, the analysis procedure with the ISOCS was optimized for quantification of radioactive waste. To this end, a standard radioactive waste drum at KEPCO NF and low-level radioactive waste drums at Korea Radioactive Waste Agency (KORAD) were measured. The performance of the ISOCS was then evaluated by Monte Carlo simulations, Multi-Group Analysis for Uranium (MGAU) code, and destructive analysis. As a result, the ISOCS showed good performance in the quantification of uranium for a drum with the homogenized simple geometry and long measurement time. It is confirmed that the ISOCS gamma spectroscopy technique could be used for control and accountancy of nuclear materials contained in a radioactive waste drum.