KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong

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.

Radioactive contaminants, such as 137Cs, are a significant concern for long-term storage of nuclear waste. Migration and retention of these contaminants in various environmental media can pose a risk to the surrounding environment. The distribution coefficient (Kd) is a critical parameter for assessing the behavior of these contaminants and can introduce significant errors in predicting migration and remediation options. Accurate prediction of Kd values is essential to assess the behavior of radioactive contaminants and to ensure environmental safety. In this study, we present machine learning models based on the Japan Atomic Energy Agency Sorption Database (JAEA-SDB) to predict Kd values for Cs in soils. We used three different machine learning models, namely the random forest (RF), artificial neural network (ANN), and convolutional neural network (CNN), to predict Kd values. The models were trained on 14 input variables from the JAEA-SDB, including factors such as Cs concentration, solid phase properties, and solution conditions which are preprocessed by normalization and log transformation. We evaluated the performance of our models using the coefficient of determination (R2) value. The RF, ANN, and CNN models achieved R2 values of over 0.97, 0.86, and 0.88, respectively. Additionally, we analyzed the variable importance of RF using out-of-bag (OOB) and CNN with an attention module. Our results showed that the initial radionuclide concentration and properties of solid phase were important variables for Kd prediction. Our machine learning models provide accurate predictions of Kd values for different soil conditions. The Kd values predicted by our models can be used to assess the behavior of radioactive contaminants in various environmental media. This can help in predicting the potential migration and retention of contaminants in soils and the selection of appropriate site remediation options. Our study provides a reliable and efficient method for predicting Kd values that can be used in environmental risk assessment and waste management.

LILW disposal repository in Gyeongju, South Korea is considered with a concrete mixture that uses Ordinary Portland Cement (OPC) partially substituted with supplementary cementitious materials (SCMs). The degradation of cementitious materials that result from chemical and physical attacks is a major concern in the safety of radioactive waste disposal. We present a reactive transport model utilized as one of the geochemical simulation approaches for the timescales of concern that range from hundreds to thousands of years. The purpose of this study is to investigate the sensitivity of parameters in concrete disposal systems and to evaluate the influence of various assumptions on the chemical degradation of the systems using a reactive transport model. A reactive transport model in the concrete disposal vault was developed to evaluate the behavior of engineered barriers composed of cementitious materials. The sensitivity analysis was performed using reactive transport models through the coupling between COMSOL and PHREEQC. The databases selected for the analysis are the Thermochimie database presented by ANDRA. Among many variables considered, two variables that can highly affect chemical degradation were selected for detailed sensitivity analysis for dealing with uncertainties. This is important because the chemical degradation mechanism is generally sensitive to precipitation and diffusion coefficient. The first factor is precipitation, which might be the most important factor in chemical degradation because it acts as a calcium leaching of cementitious materials in a disposal system in a highly alkaline environment, increasing the porosity of the system. To predict the change in annual precipitation, the measurement of the precipitation observatory station in the nearest area of Gyeongju for the past 80 years was collected. The second factor is the diffusion coefficient, which plays an essential role in the durability of the concrete disposal system, promoting the decalcification of cementitious minerals, accelerating system degradation, and increasing the porosity of its system, thereby facilitating the migration of radionuclides. The diffusion coefficient values used in studies similar to this work were calculated and evaluated using the box-and-whisker method. The results of the sensitivity analyses for the reactive transport model in the concrete disposal system will be presented. The sensitivity cases show that the results obtained are much more sensitive to changes in transport parameters.

Domestic NPPs had produced the paraffin-solidifying concentrate waste (PSCW) for nearly 20 years. At that time radioactive waste management policy of KHNP was to reduce the volume and to store safely in site. The PSCW has been identified not to meet the leaching index after introducing the treatment system. PSCW has to be treated to meet current waste acceptance criteria (WAC) for permanent disposal. PSCW consists of dried concentrate 75% and paraffin 25% of volume. When PSCW is separated into a dried concentrate and a paraffin by solubility, total volume separated is increased twice. Final disposal volume of dried concentrate can reach to several times when solidifying by cement even considering exemption. Application of polymer solidification technology is difficult because dried concentrate is hard to make form to pellet. When PSCW is packaged in High Integrity Container (HIC), volume of PSCW is equal to the volume before package. The packaging process of HIC is simple and is no necessary of large equipment. It is important to recognize that HIC was developed to replace solidification of waste. HIC has as design goal a minimum lifetime of 300 years under disposal environment. The HIC is designed to maintain its structural integrity over this period, to consider the corrosive and chemical effects of both the waste contents and the disposal environment, to have sufficient mechanical strength to withstand loads on the container and to be capable of meeting the requirements for a Type A transport Package. The Final waste form is required for facilitating handling and providing protection of personnel in relation to solidification, explosive decomposition, toxic gases, hazardous material, etc. Structural stability of final waste form is required also. Structural stability of the waste can be provided by the waste itself, solidifying or placing in HIC. Final waste form ensure that the waste does not structurally degrade and affect overall stability of the disposal site. The HIC package contained PSCW was reviewed from several points of view such as physicochemical, radiological and structural safety according to domestic WAC. The result of reviewing shows that it has not found any violation of WCP established for silo type disposal facility in Gyeongju city.

The Ag0-containing sorbents synthesized by Na, Al, and Si alkoxides have higher maximum iodine capture capacity and textural properties than zeolite-based Ag0-containing sorbents. However, these sorbents were prepared in the form of granules via a step for cutting cylindrical alcogels. Since asmade sorbents decreased packing density, they must be additionally crushed and then classified into an appropriate size for increasing packing density. The bead formation in the step of sol-gelation could bring about the simplification of sorbent preparation process and an improvement of packing density. In the Na, Al, and Si alkoxides as starting materials, sol solution was hydrophilic and lower density than vegetable oil, which transformed sol droplets to sol-gel beads. Thus, in these precursors, sol droplets, which must be sprayed by single nozzle placed at bottom side of oil column, can rise up through oil column. Acetic acid (HOAc) was used as the catalyst for the hydrolysis of Na alkoxide (TEOS) and gelation of the Na+AlSi-OH alcosol. For obtaining sol-gel beads, experiments were performed by the flowrate change of sol solution and HOAc at different nozzle sizes using soybean oil column of 1 m in length. At a sol/HOAc flowrate ratio of 3.85, some Na+AlSi-OH alcogel beads were obtained. After the Ag/Na ion-exchange, Ag content in Ag+AlSi-OH hydrogel was low due to reaction between Na+ and HOAc during sol-gelation and aging step. The Ag+AlSi-OH hydrogel with high Ag content could be prepared by Na addition. After the solvent exchange and drying at ambient pressure, the bead sorbents had higher Ag0 content and larger pore size than granular sorbents. However, further experiments are needed to increase yield rate in bead sorbent.

Nuclear fusion energy is considered as a future energy source due to its higher power density and no emission of greenhouse gas. Therefore, various researches on nuclear fusion is being conducted. One of the key materials for the nuclear fusion research is tritium because the D-T reaction is the main reaction in the nuclear fusion system. However, that tritium can also be used for non-peaceful purposes such as hydrogen bombs. Therefore, it is necessary to establish the safeguards system for tritium. In that regards, this study analyzed the possibility of applying safeguards to tritium. To achieve this objective, the tritium production capacity through the light water reactor was analyzed. Tritium Production Burnable Absorber Rod (TPBAR) was modeled through the MCNP code, and tritium production was analyzed. The TPBAR is composed of a cylindrical tube with a double coating of aluminum and zirconium, filled with a sintered lithium aluminate (LiAlO2) pellet to prevent the release of tritium. Tritium is produced by the reaction of Li-6 in the TPBAR with neutrons, and it is extracted and stored at the Tritium Extraction Facility (TEF). As a result, the tritium production increased as the burnup and Li-6 mass increased. In addition, when the tritium produced in this way was transferred to TEF and extracted through the process, the application of safeguards measures was analyzed. To this end, various safeguards measures were devised, such as setting an Material Balance Area (MBA) for TEF and analyzing Material Balance Period (MBP). As there is no designated Significant Quantity (SQ) for tritium, cases were classified based on the type and form of nuclear weapons to estimate the Sigma MUF (Material Unaccounted For) of the TEF. Finally, the comprehensive application of safeguards to tritium was discussed. This research is expected to contribute to the establishment of IAEA safeguards standards related to tritium by applying the findings to actual facilities.

Milling facilities, which belong to the front end of the nuclear fuel cycle, are essential steps for utilizing uranium in nuclear power generation. These milling facilities currently provide the International Atomic Energy Agency (IAEA) with the location and annual production capacity of the facility through the Additional Protocol (AP, INFCIRC/540) and grant IAEA inspectors on-site sampling authority to apply safeguards to the facility. However, since milling facilities process a large amount of nuclear material and the product uranium ore concentrate (UOC) is bulk material, the absence of accounting for the facility could pose a potential risk of nuclear proliferation. Therefore, this study proposes technical approach that can be utilized for safeguards in milling facilities. Since the half-life of uranium isotopes is much longer than that of its daughter, they reach the secular equilibrium condition. However, after milling process, the fresh tailing showed the break of that secular equilibrium. As time goes on, they newly reach another secular equilibrium condition. Based on this observation, this study discussed the feasibility of the ratio method in safeguards purpose. The scenario applied in this study was 1% of uranium mill tailing. It was observed that the U-238/Th-234 and U- 238/Pa-234m ratios in fresh milling tails varied as a function of time after discharging, particularly during the first one year. This change can be worked as a significant signature in terms of safeguards. In conclusion, the ratio method in mill tails could be applicable for safeguards of nuclear milling facility.

Coupled thermo-hydraulic-mechanical (THM) processes are essential for the long-term performance of deep geological disposal of high-level radioactive waste. In this study, a numerical sensitivity analysis was performed to analyze the effect of rock properties on THM responses after the execution of the heater test at the Kamaishi mine in Japan. The TOUGHFLAC simulator was applied for the numerical simulation assuming a continuum model for coupled THM analysis. The rock properties included in the sensitivity study were the Young’s modulus, permeability, thermal conductivity, and thermal expansion coefficients of crystalline rock, rock salt, and clay. The responses, i.e., temperature, water content, displacement, and stress, were measured at monitoring points in the buffer and near-field rock mass during the simulations. The thermal conductivity had an overarching impact on THM responses. The influence of Young’s modulus was evident in the mechanical behavior, whereas that of permeability was noticed through the change in the temperature and water content. The difference in the THM responses of the three rock type models implies the importance of the appropriate characterization of rock mass properties with regard to the performance assessment of the deep geological disposal of high-level radioactive waste.

Because of the small number of spacecraft available in the Earth’s magnetosphere at any given time, it is not possible to obtain direct measurements of the fundamental quantities, such as the magnetic field and plasma density, with a spatial coverage necessary for studying, global magnetospheric phenomena. In such cases, empirical as well as physics-based models are proven to be extremely valuable. This requires not only having high fidelity and high accuracy models, but also knowing the weakness and strength of such models. In this study, we assess the accuracy of the widely used Tsyganenko magnetic field models, T96, T01, and T04, by comparing the calculated magnetic field with the ones measured in-situ by the GOES satellites during geomagnetically disturbed times. We first set the baseline accuracy of the models from a data-model comparison during the intervals of geomagnetically quiet times. During quiet times, we find that all three models exhibit a systematic error of about 10% in the magnetic field magnitude, while the error in the field vector direction is on average less than 1%. We then assess the model accuracy by a data-model comparison during twelve geomagnetic storm events. We find that the errors in both the magnitude and the direction are well maintained at the quiet-time level throughout the storm phase, except during the main phase of the storms in which the largest error can reach 15% on average, and exceed well over 70% in the worst case. Interestingly, the largest error occurs not at the Dst minimum but 2–3 hours before the minimum. Finally, the T96 model has consistently underperformed compared to the other models, likely due to the lack of computation for the effects of ring current. However, the T96 and T01 models are accurate enough for most of the time except for highly disturbed periods.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

Currently, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated in Korea. For the disposal of the radioactive wastes, the transport demand is expected to increase. Prior to transportation, it is necessary to evaluate the radiation risk of transportation to confirm that is not high. In Korea, there is no transportation risk assessment code that reflects domestic characteristics. Therefore, foreign assessment codes are used. In this study, before developing the overland transportation risk assessment code that reflects domestic characteristics, we analyzed the radiation risk assessment methodology in transportation accident codes developed in other countries. RADTRAN and RISKIND codes were selected as representative overland transportation risk assessment codes. For the two codes we analyzed accident scenarios, exposure pathways, and atmospheric diffusion. In RADTRAN, the user classifies accident severity for possible accident scenarios, and the user inputs the probability for each accident severity. On the other hand, in the case of RISKIND, the accident scenarios are classified and the probabilities are determined according to the NRC modal study (LLNL, 1987) in consideration of the cask impact velocity, cask impact angle, and fire temperature. In the case of RISKIND, the accident scenarios are applied only to transportation of spent nuclear fuel, and cannot be defined for low and intermediate-level radioactive waste. However, in the case of RADTRAN, since the severity and probability of accidents are defined by user, it can be applied to low and intermediate-level radioactive wastes. As the exposure pathways considered in transportation accident, both RADTRAN and RISKIND consider external exposure (cloudshine and groundshine), and internal exposure (inhalation, resuspension inhalation and ingestion). In the case of RADTRAN, additionally, external exposure due to loss of shielding (LOS) is considered. Atmospheric diffusion calculation is essential to determine the extent to which radioactive materials are diffused. In both RADTRAN and RISKIND, atmospheric diffusion calculations are based on Gaussian diffusion model. Users must input Pasquill stability class, release height, heat release, wind speed, temperature and mixing height, etc. Additionally, RADTRAN can input weather information relatively simply by inputting only the Pasquill stability class fraction and selecting the US average weather option. This study results will be used as a basis for developing radioactive waste overland transportation risk assessment code that reflects domestic characteristics.

Cement is widely used as representative industrial material. In Korea, about 50 million tons of cement are consumed every year. In the manufacture of cement, raw materials containing NORM such as fly ash and bauxite are used. Therefore, the workers can be subjected to radiation exposure. The major exposure pathway in NORM industries is internal exposure due to inhalation of aerosol. Internal radiation dose due to aerosol inhalation varies depending on physicochemical properties of the aerosol. Therefore, the objective of this study was to investigate aerosol properties influencing inhalation dose in cement industries. In this study, aerosol properties were measured for two cement manufacturers. A particulate size distribution and concentration at various processing areas in cement manufacturing industries in Korea were analyzed using a cascade impactor. The mass density of raw materials and byproducts were measured using pycnometer. Shape of particulates was analyzed using SEM. The radioactivity concentration of Ra-226, Ra-228 for U/Th decay series was measured using HPGe. Particulate concentration by size was distributed log-normally with maximum at particle size about 7.2 μm in manufacturer A and 5.2 μm in manufacturer B. The mass density of fly ash and cement were 2.3±0.06, 3.2±0.02 g/cm3 respectively in manufacturer A. In manufacturer B, the mass density of bauxite and cement were 3.4±0.02, 2.9±0.01 g/cm3 respectively. The shape of particulates appeared as spherical shape in manufacturer A and B regardless of sampling area. Thus, a shape factor of unity could be assumed. The radioactivity concentrations of Ra-226, Ra-228 were 82±9, 82±8 Bq/kg for fly ash, and 25±4, 23±3 Bq/kg for cement in manufacturer A. In manufacturer B, the radioactivity concentrations of Ra-226, Ra-228 were 344±34, 391±32 Bq/kg for bauxite, and 122±13, 145±12 Bq/kg for cement. The radioactivity concentrations of Ra-226, Ra-228 in cement were less than raw materials such as fly ash and bauxite. It is because the dilution of the radioactivity concentration occurred during mixing with other raw materials in cement production process. This study results will be used as database for accurate dose assessment due to airborne particulate inhalation by workers in cement industries.

In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

The fuel fabrication facility has been built and is being operated by KAERI since licensing research reactor fuel fabrication in 2004. After almost 20 years of operation, outdated equipment for fabrication or inspection has been replaced by automated, digitalized ones to assure a higher quality of nuclear fuels. However, the generation of a large amount of radioactive waste is another concern for the replacement in terms of its volume and various types of it that should be categorized before disposal. The regulatory body, NSSC (Nuclear Safety and Security Commission) released a notice related to the classification of radioactive wastes, and most accessory equipment can be classified into the clearance levels, called self-disposal waste. In this study, the practice of self-disposal of metal radioactive waste is carried out to reduce its volume and downgrade its radioactivity. For metal radioactive waste, which is expected to occupy the most amount, analysis status and legal limitations were performed as follows: First, the disposal plan was established after an investigation of the use history for equipment. Second, those were classified by types of materials, and their surface radio-contamination was measured for checking self-disposable or not. After collecting data, the plan for the self-disposal was written and submitted to the Korea Institute of Nuclear Safety (KINS) for approval.

The reliable information on the hydraulic characteristics of rock mass is one of the key site factors for design and construction of deep subsurface structures such as geological radioactive nuclear waste disposal repository, underground energy storage facility, underground research laboratory, etc. In order to avoid relying on foreign field test technology in future projects, we have independently designed and made integrated type main frame, 120 bar waterproof downhole sonde, and 1,200 m wireline cable winch through a series of R&D activities. They are core apparatuses of the Deep borehole Hydraulic Test System (DHTS). Integration of individual test equipment into a single main frame allows safe and efficient work in the harsh field condition. The DHTS was developed aiming primarily for constant pressure (head) injection test and pulse test in deep impermeable rock mass. The maximum testing depth of the DHTS is about 1,050 m from the surface. Using this system, it is possible to make precise stepwise control of downhole net injection pressure in less than 2.0 kgf/cm2 with dual hydraulic volume controller and also to inject and measure the very low flow rate below 0.01 l/min with micro flow rate injection/control module. Over the past two years, we have successfully completed more than 50 in situ hydraulic tests at 5 deep boreholes located in the Mesozoic granite and sedimentary rock regions in Korea. Among them, the deepest testing depth was more than 920 m. In this paper, the major characteristics of the DHTS are introduced and also some results obtained from the high precision field tests in the deep and low permeable rock mass environment are briefly discussed.

The parent and daughter nuclides in a radioactive decay chain arrive at secular equilibrium once they have a large half-life difference. The characteristics of this equilibrium state can be used to estimate the production time of nuclear materials. In this study, a mathematical model and algorithm that can be applied to radio-chronometry using the radioactive equilibrium relationship were investigated, reviewed, and implemented. A Bateman equation that can analyze the decay of radioactive materials over time was used for the mathematical model. To obtain a differential-based solution of the Bateman equation, an algebraic numerical solution approach and two different matrix exponential functions (Moral and Levy) were implemented. The obtained result was compared with those of commonly used algorithms, such as the Chebyshev rational approximation method and WISE Uranium. The experimental analysis confirmed the similarity of the results. However, the Moral method led to an increasing calculation uncertainty once there was a branching decay, so this aspect must be improved. The time period corresponding to the production of nuclear materials or nuclear activity can be estimated using the proposed algorithm when uranium or its daughter nuclides are included in the target materials for nuclear forensics.