The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

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.

The segmentation of activated components is considered as a one of the most important processes in decommissioning. The activated components, such as reactor vessel and reactor vessel internals, are exposed to neutron from the nuclear fuel and classified to intermediate, low, and very low-level wastes. As it is expected, the components, which are closed to nuclear fuel, exhibit higher degree of specific activity. After the materials were exposed to neutrons, their original elements transform to other nuclides. The primary nuclides in activated stainless steel are 55Fe, 63,59Ni, 60Co, 54Mn, etc. The previous study indicates that the specific activity of individual nuclide is strongly depends on the material compositions and impurities of the original materials. The 59Co is the one of the most important impurities in stainless steel and carbon steel. In this paper, the relationship between individual nuclides in activation analysis of activated components was studied. The systematic study on specific activity of primary nuclides will be discussed in this paper to understand the activation tendency of the components.

Dry active waste (DAW) contains substantial amount of cellulose related materials. The DAW are usually classified as low and/or very low-level waste. In Korea, three types of disposal facilities have been considered: silo, engineering barrier, and land-fill. Currently, only the silo type disposal facility is in operation. Around 27 thousand drums were disposed in silo. Massive amount of cement concrete is used in construction of silo. The ground waste, which flow through the concrete structure, shows higher pH than as it is. It is generally known that the pH of silo is ~12.47 in Korea, when considering construction material, filling material, and property of ground water. It is expected that the cellulose in DAW will be partially transformed to isosaccharinic acid (ISA). It is generally accepted that the ISA plays a negative role in safety analysis of disposal facility by stimulation of specific nuclides. Various factors affect the degradation of cellulose containing radioactive waste, such as degree of polymerization, pH of disposal condition, interaction between concrete structure and ground water, etc. In this paper, the disposal safety analysis of cellulose containing radioactive, usually paper, cotton, wood, etc., are studied. The degradation of cellulose with respect to degree of polymerization, pH of neighboring water, filling material of silo, etc. are reviewed. Based on the review results, it is reasonable to conclude that the substantial amount of DAW could be disposed in silo.

Nuclear weapon generates huge amount of radioactive fallout which is extremely dangerous. The fallout gradually falls to the ground and then covers every surface in city and nature. A hydrogel decontamination medium has been developed to clean the surface polluted by the fallout. The hydrogel is soluble in water so the used hydrogel can be simply removed from the surface by washing. However, significant amount of waste water, containing the radioactive fallout, is generated with this process. In this respect, it is necessary to secure alternative technical options for the used hydrogel recovery. In this study, a steam-suction process was suggested for the used hydrogel recovery. Contaminated stainless steel surface, with fixed simulated fallout particles, was prepared for test. The simulated fallout particles were obtained by high-temperature treatment of a mixture of natural soil, used concrete, and Fe2O3. The hydrogel, composed of poly-vinyl alcohol and borax, was spread onto the contaminated stainless steel surface. The hydrogel was soft at first and it gradually becomes rigid with time. The used hydrogel was recovered by suction with a simultaneous steam spraying to soften the rigid gel. As a result, the clean surface of the stainless steel without the simulated fallout particles was obtained, showing the feasibility of this technique for the used hydrogel recovery.

The operation and decommissioning of nuclear power plants (NPPs) creates waste in the process of handling radioactively contaminated material, which must be disposed of in a repository or can be recovered of in the same way as conventional waste in the course of handling radioactively contaminated materials. For buildings or sites of NPPs it also has to be decided under what conditions they can continue to be used for other, conventional purposes or demolished. This decision is referred to as “release from supervision under nuclear and radiation protection law” or “clearance” in short. The clearance levels applicable in Germany according to the Radiation Protection Ordinance have been defined such that a radiation dose (hereinafter referred to as “dose”) of 10 μSv per year is not exceeded. The vast majority of the materials resulting from the dismantling of a nuclear power plant (e.g. most of the massive concrete structures) are neither contaminated nor activated. Thus, there is no need to treat these materials as radioactive waste. Emplacement of uncontaminated masses which in Germany is essentially several million tonnes of building rubble in a repository would require additional construction of such facilities, which, in view of the negligible hazard potential, from the point of view of the Nuclear Waste Management Commission (ESK) is clearly to be rejected both economically and, in particular, ecologically. Alternative ways are increasingly discussed in public, such as the abandonment of buildings after gutting, i.e. refraining from demolition of the controlled area buildings of NPPs. Also, another proposal discussed in public, the landfilling or the long-term storage of cleared material at the site, does not offer any safety-related advantages either in the view of the ESK. If, after completion of all dismantling work, the building has been decontaminated such that the clearance levels for buildings are complied with further use of the building rubble resulting from demolition is harmless from a radiological point of view. For these reasons, Germany has deliberately decided to use clearance as an essential measure in the dismantling of NPPs. If it is intended to conventionally reuse or depose of virtually contaminant-free material from controlled areas, it must first undergo a clearance procedure. The prerequisites that must be fulfilled for clearance are regulated in the Radiation Protection Ordinance, which includes two basic clearance pathways: unrestricted and specific clearance. In the following, the basic process of clearance is briefly presented and illustrated for a better understanding. It comprises five steps. Step 1-Radiological characterization by sampling, Step 2-Dismantling of plant components in the controlled area, Step 3- Decontamination, Step 4-Decission measurements, Step 5-Clearacnce and further management. The entire clearance process is governed by a clearance notice and is carried out under the supervision of the competent authority under nuclear and radiation protection law or the independent authorized expert commissioned by it. The clearance pathways contained in the Radiation Protection Ordinance have proven themselves in practice. They permit safe and proper management of material from dismantling and release of the site from supervision under nuclear and radiation protection law. These German regulatory procedures should be taken into account and deregulation and removal should be used as appropriate and necessary tools in the process of decommissioning NPPs in order to return non-hazardous materials to the material cycle or for conventional disposal.

The removal of cesium (Cs) from contaminated clay minerals is still a challenge due to the limited efficiency of the process. Thus, this study aimed to enhance the removal for Cs+ ions during the conventional acid washing process by incorporating a bead-type adsorbent. Polyacrylonitrile-based nickel potassium hexacyanoferrate (NiFC-PAN) was utilized as the Cs adsorbent to selectively adsorb Cs+ ions in a strongly acidic solution that contained competing ions. To enable easy separation of clay particles and protect the adsorbent from harsh environmental conditions, PAN was deliberately constructed as large beads. The synthesized adsorbent (NiFC/PAN in a 2:1 ratio) displayed high selectivity for Cs+ ions and had a maximum capacity of 162.78 mg/g for Cs+ adsorption in 0.5 M HNO3 solution. Since NiFC-PAN exhibited greater Cs selectivity than the clay mineral (hydrobiotite, HBT), adding NiFC-PAN during the acid washing substantially increased Cs desorption (73.3%) by preventing the re-adsorption for Cs+ ions on the HBT. The acid treatment in the presence of NiFCPAN also significantly decreased the radioactivity of 137Cs-HBT from 209 to 27 Bq/g, resulting in a desorption efficiency of 87.1%. Therefore, these findings suggest that the proposed technique is a potentially useful and effective method for decontaminating radioactive clay.

Metakaolin-based geopolymers have shown promise as suitable candidates for 14C immobilization and final disposal. It has been shown that the physicochemical properties of metakaolin wasteforms meet, and often far exceeding, the strict compression strength and leaching acceptance criteria of the South Korea radioactive waste disposal site. However, it is not possible to analyze and characterize the internal structure of the geopolymer wasteform by conventional characterization techniques such as microscopy without destruction of the wasteform; an impractical solution for inspecting wasteforms destined for final disposal. Internal inspection is important for ensuring wastes are homogenously mixed throughout the wasteform and that the wasteform itself does not pose any significant defects that may have formed either during formulation and curing or as a result of testing prior to final disposal. X-ray Computed Tomography (XCT) enables Non-Destructive Evaluation (NDE) of objects, such as final wasteforms, allowing for both their internal and external, characterization without destruction. However, for accurate quantification of an objects dimensions the spatial resolution (length and volume measures) must be know to a high degree of precision and accuracy. This often requires extensive knowledge of the equipment being used, its precise set-up, maintenance and calibration, as well as expert operation to yield the best results. A spatial resolution target consists of manufactured defects of uniformed dimensions and geometries which can be measured to a high degree of accuracy. Implementing the use of a spatial resolution target, the dimensions of which are known and certified independently, would allow for rapid dimensional calibration of XCT systems for the purpose of object metrology. However, for a spatial resolution target to be practical it should be made of the same material as the intended specimen, or at least exhibit comparable X-ray attenuation. In this study, attempts have been made to manufacture spatial resolution targets using geopolymer, silica glass, and alumina rods, as well as 3D printed materials with varying degrees of success. The metakaolin was activated by an alkaline activator KOH to from a geopolymer paste that was moulded into a cylinder (Diameter approx. 25 mm). The solidified geopolymer cylinder as well as both the silica glass rod and alumina rod (Diameter approx. 25 mm) we cut to approximately 4 mm ± 0.5 mm height with additional end caps cut measuring 17.5 mm ± 2.5 mm height. All parts were then polished to a high finish and visually inspected for their suitability as spatial resolution targets.

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.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

Bentonite is a widely used buffer material in high-level radioactive waste repositories due to its favorable properties, including its ability to swell and low permeability. Bentonite buffers play an important role in safe disposal by providing a low permeability barrier and preventing radionuclides migration into the surrounding rock. However, the long-term performance of the bentonite buffer is still an area of research, and one of the main concerns is the erosion of the buffer due to swelling and groundwater flow. Erosion of the bentonite buffer can have a significant impact on repository safety by reducing the integrity of the buffer and forming colloids that can transport radionuclides through groundwater, potentially increasing the risk of radionuclide migration. Therefore, understanding the mechanisms and factors that influence the erosion of the bentonite buffer is critical to the safety assessment of high-level radioactive waste repositories. In this study, we attempted to develop the bentonite buffer erosion model using Adaptive Processbased total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). First, the erosion phenomenon was divided into two stages: bentonite buffer penetration into rock fractures and colloid formation. As an initial step in the development of the buffer erosion model, a bentonite buffer intrusion into the fracture and consequent degradation of buffer property were considered. For this purpose, a tworegion model based on the dynamic bentonite diffusion model was adopted which is one of the methods for simulating bentonite buffer intrusion. And, it was assumed that the buffer properties, such as density, porosity and permeability, thermal conductivity, modulus of elasticity, and mechanical strength, are degraded as the buffer erodes. The bentonite buffer degradation model developed in this study will serve as a foundation for the comprehensive buffer erosion model, in conjunction with the colloidal formation model in the future.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

It is important to make a strategy for clearance-level radioactive waste. Sampling and disposal plans should be drawn up with characteristics of target waste. In this paper, a target clearance-level radioactive waste is used in a laboratory for experiments with Cs-137 and Co-60, unsealed radioactive sources with gamma radiation isotopes. Therefore, it is enough to analyze with HPGe to check the contaminant level. The laboratory fume hood combined multiple materials, which means some are volume contamination and others are surface contamination. The wood, plastic, and drywall boards, which are absorbent volume contaminated parts and make up PVC pipes, base cabinet doors, backside baffles, etc., will be sampled with coring methods. The metals and glasses, which are unabsorbent, surface-contaminated parts, are sampled with smear methods. The work surface, baffles, exhaust plenum, and glass sash inside parts have a high possibility of being contaminated. The hood body, flame, base cabinet, PVC pipe (the rare end of the filter), and blower transition case have a low possibility of becoming contaminated. When we checked with HPGe, except for the work surface (which was below clearance level), other parts were less than MDA. The highest radionuclide concentration was in PVC pipe: Cs-137C 3.91E-02 (Bq/g), Co-60 4.54E- 03 (Bq/g). It is less than clearance level. Therefore, the waste was applied for the clearance level radioactive wastes and got permission from the regulatory body.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.

Level measurement of liquid radwaste is essential for inventory management of treatment system. Among various methods, level measurement based on differential pressure has many advantages. First, it is possible to measure the liquid level of the system regardless of liquid type. Second, as the instrument doesn’t need to be installed near the tank, there is no need to contact the tank when managing it. Therefore, workers’ radiation dose from the system can be decreased. Finally, although it depends on the accuracy, the price of the instrument is relatively low. With these advantages, in general, liquid radwaste level in a tank is measured using differential pressure in the treatment system. Not only the advantages described above, there are some disadvantages. As the liquid in the system is waste, it is not pure but has some suspended materials. These materials can be accumulated in tanks and pipes where the liquids move to come into direct contact with pneumatic pipes that are essential in differential pressure instruments. As a result, in case of a treatment using heat source, the accumulated materials may become sludge causing interference in pneumatic pipes. And this can change the pressure which also affects the level measured. In conclusion, in case of liquid storage tanks in which the situation cannot be checked, the proficiency of an operator becomes important.

Surface water temperature of a bay (from the south to the north) increases in spring and summer, but decreases in autumn and winter. Due to shallow water depth, freshwater outflow, and weak current, the water temperature in the central to northern part of the bay is greatly affected by the land coast and air temperature, with large fluctuations. Water temperature variations are large in the north-east coast of the bay, but small in the south-west coast. The difference between water temperature and air temperature is greater in winter and in the south-central part of the bay than that in the north to the eastern coast of the bay where sea dykes are located. As the bay goes from south to north, the range of water temperature fluctuation and the phase show increases. When fresh water is released from the sea dike, the surrounding water temperature decreases and then rises, or rises and then falls. The first mode of empirical orthogonal function (EOF) represents seasonal variation of water temperature. The second mode represents the variability of water temperature gradient in east-west and north-south directions of the bay. In the first mode, the maximum and the minimum are shown in autumn and summer, respectively, consistent with seasonal distribution of surface water temperature variance. In the second mode, phases of the coast of Seosan~Boryeong and the east coast of Anmyeon Island are opposite to each other, bordering the center of the deep bay. Periodic fluctuation of the first mode time coefficient dominates in the one-day and half-day cycle. Its daily fluctuation pattern is similar to air temperature variation. Sea conditions and topographical characteristics excluding air temperature are factors contributing to the variation of the second mode time coefficient.

Some of the metal waste generated from KEPCO NF is being disposed of in the form of ingots. An ingot is a metal that is melted once and then poured into a mold to harden, and it is characterized by a uniform distribution of radioactive material. When measuring the uranium radioactivity in metal ingot with HPGe detector, 185.7 keV of U-235 is used typically because most gamma rays emitted at U-235 are distributed in low-energy regions below 200 keV. To analyze radioactivity concentration of U-235 with HPGe detector more accurately, self-attenuation due to geometrical differences between the calibration source and the sample must be corrected. In this study, the MCNP code was used to simulate the HPGe gamma spectroscopy system, and various processes were performed to prove the correlation with the actual values. First an metal ingottype standard source was manufactured for efficiency calibration, and the GEB coefficient was derived using Origin program. And through the comparison of actual measurements and simulations, the thickness of the detector’s dead layers were defined in all directions of Ge crystal. Additionally instead of making an metal ingot-type standard source every time, we analyzed the measurement tendency between commercially available HPGe calibration source (Marinelli beaker type) and the sample (metal ingot type), and derived the correction factor for geometry differences. Lastly the correction factor was taken into consideration when obtaining the uranium radioactivity concentration in the metal ingot with HPGe gamma spectroscopy. In conclusion, the U-235 radioactivity in metal ingot was underestimated about 25% of content due to the self-attenuation. Therefore it is reasonable to reflect this correction factor in the calculation of U-235 radioactivity concentration.

The engineered barrier system (EBS) for deep geological disposal of high-level radioactive waste requires a buffer material that can prevent groundwater infiltration, protect the canister, dissipate decay heat effectively, and delay the transport of radioactive materials. To meet those stringent performance criteria, the buffer material is prepared as a compacted block with high-density using various press methods. However, crack and degradation induced by stress relaxation and moisture changes in the compacted bentonite blocks, which are manufactured according to the geometry of the disposal hole, can critically affect the performance of the buffer. Therefore, it is imperative to develop an adequate method for quality assessment of the compacted buffer block. Recently, several non-destructive testing methods, including elastic wave measurement technology, have been attempted to evaluate the quality and aging of various construction materials. In this study, we have evaluated the compressive wave velocity of compacted bentonite blocks via the ultrasonic velocity method (UVM) and free-free resonant column method (FFRC), and analyzed the relationship among compressive wave velocity, dry density, thermal conductivity, and strength parameter. We prepared compacted bentonite block specimens using the cold isostatic pressure (CIP) method under different water content and CIP pressure conditions. Based on multiple regression analysis, we suggest a prediction model for dry density in terms of manufacturing conditions. Additionally, we propose an empirical model to predict thermal conductivity and unconfined compressive strength based on compressive wave velocity. The database and suggested models in this study can contribute to the development of quality assessment and prediction techniques for compacted buffer blocks used in the construction of a disposal repository.

To prove the long-term safety of deep geological repository, the safety assessment is needed to ensure that the expected performance of repository satisfies the regulatory standards. Scenario development is process of analyzing events and evolutions that can directly or indirectly affect the performance of a disposal system and is a pre-step for quantitative safety assessment. Scenarios are used to identify and define cases to be assessed by numerical modeling, and cases are mainly divided into normal (also called the ‘reference’ and ‘expected evolution’) and abnormal scenarios. Mainly two approaches have been used to set up scenarios. One is a bottom-up approach that starts with features, events and processes (FEPs). This approach can analyze the evolution and events related to the performance of the disposal system in an inductive manner. The other is top-down approach that analyzes the events and evolution of disposal system, focusing on situations that may affect the safety function of the components. This approach starts with a set of intuitively predefined expected failures of safety function. Combining the two approaches is more effective in demonstrating comprehensiveness which is a main challenge of scenario analysis, and almost national radioactive waste management institutions combine top-down and bottom-up approaches for development of scenarios. An approach combining the two approaches is called a hybrid approach, and the detailed method differs from each institution and has not been determined. In this study, some work for constructing the scenario using hybrid approach was performed. Firstly, defining each component’s safety function and screening FEPs according to several rules were performed for a generic repository. Secondly, we extracted performance factors that are considered likely to affect safety functions. And lastly, we integrated FEPs correlated with performance factor to simplify the analysis. These results will be material to construct the scenario using hybrid approach.