To ensure the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

In order to apply indirect methods (such as scaling factors) to assess the radionuclide inventory of waste generated by nuclear power plants, it is essential to first evaluate the correlation coefficient between key radionuclides and those that are difficult to measure (DTM). The benchmark for the correlation coefficient (r) applied in indirect assessments is set at 0.6, and its significance can vary based on both its value and the size of the dataset. For instance, deriving a correlation coefficient using three data points versus utilizing a dataset with a hundred data points would yield different implications. This study addresses the variance in correlation coefficients based on data selection and presents a methodology for validating the significance of these coefficients. Additionally, we will discuss how these variances may impact the results of indirect assessments, such as scaling factor evaluations.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

Deep disposal facility for High-Level radioactive Waste (HLW) uses a multi-barrier system to prevent the leakage of radionuclide. As a part of the engineered barrier, bentonite is primarily considered as the main buffering material. This is due to the adsorption and swelling properties of the bentonite, which are expected to effectively impede leakage of the radionuclide. In many cases, adsorption is generally regarded as occurring only within the buffer zone. However, several research has been conducted to explore the possibility of bentonite intrusion into the Excavation- Damaged Zone (EDZ) generated during excavation processes, because of the swelling properties of the bentonite. Generally, for host rock near the deep disposal facility such as granite, groundwater flows through the fracture network. Therefore, analysis of the characteristics of the fracture network is essential for predicting the behavior of radionuclide in groundwater. Accordingly, the bentonite intrusion into the fracture network is critical for safety assessment of the deep disposal facility. To analyze this, hydro-geochemical model was established utilizing COMSOL Multiphysics and PHREEQC, observing changes of the behavior of U (VI) along fracture network due to the swelling of bentonite. Modeling was conducted with progressively changing intrusion depth of the bentonite. According to the results, the behavior of U (VI) exhibited significant changes depending on the connectivity of the fractures. Based on the distribution characteristics of the fracture network, heterogeneous groundwater flow was observed. U (VI) was transported through the preferential pathway, which indicates high connectivity, due to the rapid groundwater flow. Notably, when changing the intrusion depth of bentonite, significant differences in behavior of U (VI) were observed in the 0-20 cm case. In contrast, as the intrusion depth increased, it was observed that differences became less evident. These results indicate that changes in the properties of fracture network in EDZ due to the swelling of bentonite significantly influence the behavior of U (VI).

Safety assessments for geological disposal systems extend over tens of thousands of years, taking into account the radiotoxicity decay period of spent nuclear fuel. During this extensive period, the biosphere experiences multiple glacial cycles, and fluctuations in seawater amounts, attributed to the formation and melting of glaciers, lead to global sea level changes known as eustacy. These sea level changes can directly influence the land-sea interface and groundwater flow dynamics, consequently affecting the pathways of radionuclide transport - an essential element of dose assessment. Therefore, this study aims to investigate how glacial cycles and sea level changes impact radionuclide transport within geological disposal systems, especially in the biosphere. To achieve this objective, we obtained climate evolution data including sea level changes for the Korean Peninsula over a 200,000-years, simulated by a General Circulation Model (GCM). These data were then employed to predict site and hydrology evolutions. The study site was conceptualized biosphere of Artificial Disposal System (ADioS), and we utilized the Soil and Water Assessment Tool (SWAT) to simulate hydrological evolution. These datasets, encompassing climate, site, and hydrology evolution, were collectively employed as inputs for the biosphere module of Adaptive Process-Based Total System Performance Assessment Framework (APro). Subsequently, the APro’s biosphere module calculated radionuclide transport in groundwater flow and its release into surface water bodies, considering the influences of glacial cycles and sea level changes. The results show that hydrologic changes due to sea level change are relatively minor, while the impact of sea level change on groundwater flow and discharge is significant. Additionally, we identified that among the water bodies within ADioS, including rivers, lakes, and oceans, the ocean exhibits the most substantial radionuclide outflow throughout the entire period. The spatiotemporal distributions of radionuclides computed within APro will be further processed into a grid format and used as input for the dose assessment module. Through this study, it was possible to determine the impact of long-term glacial cycles and sea level changes on radionuclide transport. Additionally, this module can serve as a valuable tool for providing the spatiotemporal variability of radionuclides required for enhanced dose assessments.

The increasing accumulation of spent nuclear fuel has raised interest in High-Level Waste (HLW) repositories. For example, Sweden is under construction of the KBS-3 repository. To ensure the safety of such HLW repository, various countries have been developing assessment models. In the Republic of Korea, the Korea Atomic Energy Research Institute has been developing on the AKRS model. However, traditional safety assessment models have not considered the fracture growth in the far-field host rock as a function of time. As repository safety assessments guarantee safety for million years, sustained stress naturally leads to the progressive growth of fractures as time goes on. Therefore, it becomes essential to account for fracture growth in the surrounding host rock. To address this, our study proposes a new coupling scheme between the Fracture growth model and the radionuclide transport model. That coupling scheme consists of the Cubic Law model as a fracture growth function and the GoldSim code which is a commercial software for radionuclide transport calculations. The model that adopting such fracture growth functions showed an increase of up to 15% in the release of radionuclide compared to traditional assessment models. our observations indicated that crack growth as a function of time led to an increase in hydraulic conductivity that allowed more radionuclide transport. Notably, these findings show the significance of adopting fracture growth models as a critical element in evaluating the safety of nuclear waste repositories.

The final disposal of Spent Nuclear Fuel (SNF) will take place in a deep geological repository. The metal canister surrounding the SNF is made of cast iron and copper, designed to provide longterm containment of radionuclides. Canister is intended to be safeguarded by a multiple-barrier disposal system comprising engineered and natural barriers. Colloids and gases are mediators that can accelerate radionuclide migration and influence radionuclide behavior when radionuclides leak from the canister at the end of its service life. It is very important to consider these factors in the assessment of the long-term stability of deep dispoal repository. An experimental setup was designed to observe the acceleration of nuclide behavior due to gas-mediated transport in a simulated environment with specific temperature and pressure conditions, similar to those of a deep disposal repository. In this study, experiments were conducted to simulate gas flow within an engineered barrier under conditions reflecting 1000 years post repository closure. The experiment utilized bentonite WRK with a dry density of 1.61 g/cm³ after compaction. The compacted bentonite was subsequently saturated under a water pressure of 5 MPa, equivalent to the hydrostatic pressure found 500 meters underground. Gas was introduced into the saturated bentonite at different pressures to assess the permeation behavior of the bentonite relative to gas pressure variations. Consequently, it was observed that under specific pressures, gas permeated the saturated bentonite, ascending in the form of bubbles. Furthermore, it was noted that when a continuous flow was initiated within the bentonite, erosion took place, leading to the buoyant transportation of eroded particles upward by the bubbles. The particles transported by the bubbles had a relatively small particle size distribution, and cesium also tended to be transported by the bubbles and moved upward. When high-pressure gas is generated at the interface of the canister and the buffer, flow through the buffer can occur, and cationic nuclides such as cesium and strontium can be attached to the gas bubble and migrate. However, the pressure of the gas to break through the saturated buffer is very high, and the amount of cesium transported by the gas bubbles is very limited.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

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

Support for nuclear power plant (NPP) dismantling & decommissioing (D&D) industry is necessary through development of the infrastructure and the D&D technology. Because KORI#1 and Wolsong#1 is planned to decommission until around 2030. Korea research institute of decommissioing (KRID) was established through the preliminary feasibility study. KRID has plan to support nuclear companies to join D&D industry. Normal facilities (Lv.1) of KRID infracstucture are currently being constructed and radiation management facilities (Lv.2) construction is expected to begin in October. Further, KRID is planning the construction of equipment to develop the procedure for radionuclide analysis through R&D project. A total period of the R&D project is 45 months, and the total R&D funding for this period is 19.4 billion won. The ultimate goal of the R&D project is to build the infractstucture base to analyze decommissioning radioactive wastes. Furthermore, the R&D project is important to reliably perform the NPP D&D.

The ability to both assay the presence of, and to selectively remove ions in a solution is an important tool for waste water treatment in many industrial sectors, especially the nuclear industry. Nuclear waste streams contain high concentrations of heavy metals ions and radionuclides, which are extremely toxic and harmful to the environment, wildlife and humans. For the UK nuclear industry alone, it is estimated that there will be 4.9 million metric tonnes of radioactive waste by 2125, which contains a significant number of toxic radionuclides and heavy metals. This is exacerbated further by increased international growth of nuclear new build and decommissioning. Efforts to remove radionuclides have been focused on the development and optimisation of current separation and sequestering techniques as well as new technologies. Due to the large volumes of waste the techniques must be economical, simple to use and highly efficient in application. Magnetic nanoparticles (MNPs) offer a powerful enhancement of normal ion exchange materials in that they can be navigated to specific places using external magnetic fields and hence can be used to investigate challenges such as, pipework in preparation of decommissioning projects. They also have the potential to be fine-tuned to extract a variety of other radionuclides and toxic heavy metals. It has been demonstrated that with the right functional groups these particles become very strongly selective to radionuclides, such as Uranium. However, this new technology also has the potential to effectively aid nuclear waste remediation at a low cost for the separation of both radionuclides and heavy metals. In this work, we investigate the origin of the selectivity of superparamagnetic iron oxide nanoparticles (SPIONs) to Uranium by making systematic changes to the existing surface chemistry and determining how these changes influence the selectivity. Identifying the mechanism by which selected common nuclear related metals, such as Na(I), K(I), Cs(I), Ca(II), Cu(II), Co(II), Ni(II), Cd(II), Mg(II), Sr(II), Pb(II), Al(III), Mn(II), Eu(III) and Fe(III), are sorbed will allow for specific NP-target (nanoparticle) ion interactions to be revealed. Ultimately this understanding will provide guidance in the design of new targeted NP-ligand constructs for other environmental systems.

In underground repository environments, various types of engineered barriers are installed to hinder the mobility of radionuclides. Cement admixtures, especially used to improve workability for concrete, are composed of fairly high organic molecules and have a dispersing effect through bonding with the C-S-H of the concrete. Previous studies have shown that complex-forming organics like EDTA, NTA, and ISA have a significant effect on the mobility of radionuclides, but the studies on the behavior and stability of combined complexes in hydrated cement are lacking. So, we selected a commonly used polycarboxylic-ester (PCE) type cement admixture and stable Co as a surrogate of Co-60 to perform desorption experiments from hydrated cement containing the admixture. Radioactive Co is known to be a common contaminant in nuclear fission and medical facilities and considered to exist as a relatively stable phase in repositories. In addition, the evaluation of cobalt can be a standard of safety issue for other radionuclides with the presence of cement admixture in repository. In this study, cement samples were prepared at water/cement ratio of 0.55 and cured for 28 days at 23-25°C and at least 80% of humidity with varying cement admixtures of 0.0, 0.1, and 2.0wt%. To evaluate the stability of cobalt in the weathered cement, a 0.001 M HCl solution was used to simulate cement weathering conditions on a hot plate at 60°C for 1 day using a solid/liquid ratio of 1:100. Degree of weathering was confirmed using XRD analysis. The adsorption experiments were performed by adding 0.0042 mmol of cobalt (CoCl2, Sigma-Aldrich, anhydrous ≥ 98.0%) to the weathered cement for 3 days using a platform shaker at 200 rpm, and the supernatant was separated using a syringe filter (<0.20 um) before ICP-MS analysis to determine the amount of Co adsorption. Cobalt desorption was tested for the Co-adsorbed cement using 0.019 mmol of calcium (Ca(NO3)2·4H2O, Sigma-Aldrich, 99%) for 3 hours to 14 days. The results showed that adsorbed cobalt with and without cement admixture was stably bound to cement, and did not increase any noticeable Co release by 2.0wt% PCE admixture. However, additional experiments using varying contents of PCE and other admixtures should be conducted to provide a standard for assessing the safety of cement admixtures in repositories.