Measurement of the physical properties of high-temperature molten salts is important for the efficient design and operation of molten salt reactors (MSR) in which the reactor coolant and nuclear fuel are in a homogeneous liquid state. Although some crucial physical properties such as viscosity, thermal conductivity, density, etc., have been drawing much attention, relative data, especially for molten chloride salts, are scarce. Thus, it is urgent to prepare the viscosity data as one of the key transport properties in thermal hydraulics analysis. However, it is not an easy task to measure the molten salt viscosity with high accuracy due to end effect, a small gap between the chamber and spindle, thermal expansion of the chamber and spindle at high temperatures in a rotational viscometer. Additionally, molten salt temperatures inside furnace are not uniform due to the large temperature gradient inside the chamber, and therefore the assumption of laminar condition can be violated. In this study, geometric factors, which can be a major interference in the torque measurement, were considered for the accurate determination of the viscosity. We established a high-temperature molten salt viscosity measurement system with Brookfield rotational viscometer. KNO3 molten salt was used as a model substance at a temperature range of 650–773 K. In-house designed spindles and chambers were made of corrosion-resistant alumina. Thermal expansion has a significant influence on the size and shape of the chamber and spindle. The effect of thermal expansion on the conventional correction method was examined with temperature variation and distribution. Gap size variation was also investigated in order to improve the accuracy.

According to the article 18 of NSSC notice “Regulations on the delivery of low-and intermediatelevel radioactive wastes”, the consignor shall establish and implement the quality assurance program about waste management to ensure conformity with the criteria set forth in the regulations and detailed criteria proposed by the disposal facility operator, including matters related to characterization of the waste concerned. To meet the above requirement, commercially available laboratory information management system, STARLIMS from Abbott Informatics was introduced in the late of 2019 and was customized to our standardized test method in 2020. In that time, Electronic Lab Notebook (ELN), which is an electronic system to create, store, retrieve, and share fully electronic records, was tailored to replace paper lab notebook. Scientific Data Management System (SDMS), which is computer system used to capture, centrally store, catalog, and manage data, was installed. Due to the parsing ability of SDMS, human error like mistake while data entry was reduced by extracting data from measurement sheet and exporting measurement data to designated area of ELN and this feature made work efficiency improved. Afterward, validation of STARLIMS was conducted following the procedure of user acceptance testing including Operational Qualification and Performance Qualification. As a result of these activities, STARLIMS has been officially operated and applied to means to manage test results since 2021. In 2021, for user-friendly environment, updating STARLIMS was also conducted by applying SDMS to import data from other radiometric measurements including gas proportional counter (GPC), liquid scintillation counter (LSC), and low-energy Ge detector (LEGe) besides HPGe detector for gamma measurement. From implementation to operation, it is confirmed that STARLIMS has been providing reliable and stable platforms to manage laboratory information regarding measurement records and playing a significant role in tool to meet the quality assurance required.

Density of chloride molten salts is an essential physical property in the reactor core design and thermal-hydraulic design simulation, especially in molten salt reactor (MSR) design currently under development in Korea. NaCl-MgCl2-UCl3 pseudo-ternary system is one of the various candidate chloride-based salt mixtures because it has relatively-low melting point, very low vapor pressure, high thermal conductivity, etc. However, to the best of our knowledge, the density data of NaCl-MgCl2- UCl3 have not yet been measured or published worldwide, and therefore the ballpark figures of the density should be given for the preliminary reactor design. In our present study, the density estimation of NaCl-MgCl2-UCl3 based on the pseudo-binary data, i.e., NaCl-MgCl2, MgCl2-UCl3, and NaCl- UCl3, reported in the literature previously were performed using the Redlich-Kister model. Binary interaction parameter for MgCl2-UCl3 was higher than that for NaCl-MgCl2 and lower than that for NaCl-UCl3. As an example, calculated density of 0.62 NaCl: 0.18 MgCl2: 0.20 UCl3 at 873 K was 2.578 g·cm−3. In our further study, the methodology using Redlich-Kister model will be applied to more complex multicomponent systems and to other physical properties such as viscosity, thermal conductivity, surface tension, etc.

Korea Institute of Radiological and Medical Sciences provides proton irradiation service of up to 40 MeV using cyclotron. The use of such a cyclotron was approved in advance to satisfy the Nuclear Safety Act, and radiation safety was evaluated in this process. The Monte Carlo method is generally used to evaluate the shielding safety of high-energy accelerators, and MCNP 6.2 was used in the previous evaluation. In this study, in order to verify the results of previous evaluation, the calculation results of MCNP 6.2 and Particle and Heavy Ion Transport code System (PHITS) 3.24 are compared. PHITS is a general-purpose Monte Carlo particle transport simulation code that is used in many studies in the fields of accelerator technology, radiotherapy, space radiation, etc. In the previous evaluation, the effective dose by neutrons and photons generated by the collision of 40 MeV 20 μA of protons with a 10.5 mm thick beryllium target was evaluated, and in this study, this was reproduced with PHITS. As the radiation exposure evaluation for the user or pubic is evaluated based on the radiation dose and energy distribution generated around the target, the effective dose and energy distribution received by the water phantom with a radius of 1 cm on the front, side, and back of the target were calculated. T-Track, a tally of PHITS, was used to calculate effective dose, which is similar to F4 tally of MCNP 6.2 using a dose conversion factor. For the dose conversion factor, the value suggested as AP irradiation in Publication 103 was used. As a result of the calculation, the effective dose by neutrons at the front, side and back of the target was 1.42×105, 2.09×104, and 1.39×104 mSv·h−1, respectively, which was similar to 2.00×105, 1.84×104, and 2.59×104 calculated using F4 tally in MCNP. Moreover, the results of calculating the effective dose by photons using PHITS were 4.81×10, 3.10×10, and 2.66×10, respectively, and the results of calculating MCNP were 4.49×102, 6.45×10, and 9.64×10. The average energies of neutrons were 11.2, 0.69, and 0.31 MeV when calculated by PHITS, respectively, and 13.8, 7.8, and 4.6 when calculated by MCNP. Moreover, the average energies of photons were 1.98, 0.98, and 0.86 when calculated by PHITS, respectively, and 3.9, 3.2, and 2.6 when calculated by MCNP.

The radwaste facility management team is preparing for clearance of 4 MCAs in The Radwaste Form Test Facility (RFTF). The targeted waste was used for clearance level radioactive waste sample analysis and has been used for this purpose since the early 2000s. Due to the characteristics of clearance level radioactive waste, the concentration of radioactivity is very low and MCA is used with Marinelli beakers the possibility of contamination is low. Moreover, the radiation detector should not be contaminated with radioactive materials, it should be less than the clearance level. These detectors were considered surface contamination materials. To detect the contaminated spot of each material, we scanned the whole surface of a material with a gamma survey meter. After that, we took a sample from 1×1 m2 and a total of 30 samples from each MCA. The wiped filter paper was analyzed with alpha, beta low background counting systems. The results of the analysis of the smear sample of total alpha and beta nuclide radioactivity were less than MDA (α: 2.88×10−5 Bq·cm−2, β: 3.07×10−5 Bq·cm−2). The major nuclide in this facility is Co-60 and Cs-137 therefore we analyzed gamma nuclide activity with HPGe. The maximum specific activity was Co-60: 2.31×10−5 Bq·cm−2, Cs-137: 1.96×10−6 Bq·cm−2. If it is satisfied with the clearance criteria, detectors will be reused at the Radioactive Waste Treatment Facility (RWTF) room # 7251 uncontrolled area.

Radioactive materials emitted from nuclear accident or decommissioning cause soil contamination over wide areas. In the event of such a wide area of contaminated soil, decontamination is inevitable for residents to reside and reuse as industrial land. There are many ways to decontaminate these contaminated soils, but in urgent situations, the soil washing, which has a short process period and relatively high decontamination efficiency, is considered the most suitable. However, the soil washing process of removing fine soil and cesium by using washing liquid as water and adding a flocculating agent (J-AF) generates slurry/sludge-type secondary waste (Cs-contaminated soil + flocculating agent). Since this form of sludge contaminants cannot be disposed, solidification is needed using an appropriate solidification agent to treat wastes for disposal. Therefore, this study devised a treatment method of contaminated fine soils occurring after the soil washing process. This investigation prepared the simulated wastes of contaminated fine soils generated after the soil washing, and pelletized the samples using a roll compactor under the optimum operating conditions. The optimum conditions of the device were determined in the pre-test. Roll speed, feeding rate, and hydraulic pressure were 1.5 rpm, 25 rpm, and 28.44 MPa, respectively. The waste forms were manufactured by incorporating created pellets (H 6.5 × W 9.4 mm) using polymers as solidification agents. Used polymers were main ingredient (YD-128), hardener (G-1034), and diluent (LGE). The optimum mixing ratio was YD-128 : G-1034 = 65 : 35 phr, and LGE was added in an amount of 10wt% of the total mixture. To confirm the disposal suitability of the manufactured waste forms, characterization evaluation was carried out (compressive strength, thermal cycling, immersion, and leaching test). Characterization evaluation revealed a minimum compressive strength of 23.1 MPa, far exceeding 3.44 MPa of the disposal facility waste acceptance criteria. Compressive strength increased to the highest value of 31.90 MPa after immersion test. To examine leaching characteristics, the pH, Electrical Conductivity (EC) and leachability index (􀜮􀯜) of leachates were identified. As results, pH and EC consistently increased or remained constant with leaching time. The average of Co, Cs and Sr nuclides was 17.76, 17.38 and 14.04, respectively, exceeding the value of 6 in the waste acceptance criteria. Effective waste treatment/ disposal can be achieved without increasing volumes of sludge/slurry by enhancing the technique of this research by performing additional studies in the future.

According to the Atomic Energy Act of Korea, radioactive waste can be cleared when it meets the criteria, less than 10 uSv·y−1 for individual dose and 1 person · Sv·y−1 for collective dose. Consequently, it is necessary to evaluate radiation dose to get permission for regulatory clearance from the regulatory body of Korea. Several computational programs can be used for dose calculation depending on disposal methods such as landfill, incineration, and recycling. As for incineration, the effects of radionuclide emitted during combusting radwaste have to be considered to figure out exposure dose. In this study, GASPAR code is described to assess exposure dose from effluents released to the atmosphere during incinerating combustible radioactive wastes for regulatory clearance. GASPAR is the code programmed by Radiation Safety Information Computational Center at Oak Ridge National Laboratory for computing annual dose due to radioactive effluents released from a nuclear power plant to the atmosphere during routine operation. The calculating methods of the code are based on the mathematical model of U.S. NRC regulatory guide 1.109, about beta and gamma radiation from noble gas in semi-infinite plume, radioiodine, and particulates. GASPAR evaluates both individual dose and population dose. The considering pathways are composed of external exposure by plume and ground deposition of effluents, and internal exposure as a result of inhalation and food ingestion. Since the calculation model of GASPAR requires various variables about the radionuclide and disposal site, the accuracy of the results is decided by inputted values. The program contains the default values to parameters such as the humidity, fraction of deposition, and storage time of foods. However, to get permission, it is important to use the appropriate data representing the condition of the combustion scenario as substitutes for the default since the values are localized to the country where the code was developed. Therefore, dose assessment by GASPAR code can be applied for regulatory clearance by incineration, when reliable values depending on the disposal plan inputted.

Since it takes hundreds of thousands of years for the radiotoxicity of spent nuclear fuel to decrease to natural levels, interactions between each repository barrier, climate change, and geological evolutions are inevitable. These processes should be defined as the long-term evolution FEPs and considered in the performance assessment to ensure the long-term safety of the disposal system. The literature survey on geological characteristics and history of the Korean peninsula was conducted, and the list of A-KRS-FEPs which are directly or indirectly related to long-term evolutions was identified in this study. The ice age and geological change are the capital phenomena considered in the exceedingly long-term evolution before/after climate change. The historical data on ice sheets and permafrost were analyzed to investigate the effects of the ice ages on the Korean peninsula. The sealevel changes were investigated based on the research on the coastal terrace to identify the impact on uplift and shoreline change accompanying the ice age. Also, the survey on the geological history data was conducted from the perspective of tectonic activity, metamorphism, igneous activity, and seismic activities to consider the geodynamic evolution of the Korean peninsula. As results, it was suggested that 14 FEPs were directly related to climate change, 18 FEPs were directly related to geological evolution, and 47 FEPs were indirectly relevant to long-term geodynamics. The consent-based FEPs and scenarios for the long-term evolution will be developed shortly, including most of the critical long-term evolution phenomena defined in this study and which are highly probable in domestic disposal conditions. The evaluation and verification of the APro system for long-term safety will accomplish using these FEPs and scenarios.

Gases such as hydrogen can generate from the disposal canister in high-level radioactive waste disposal systems owing to the corrosion of cooper container in anoxic conditions. The gas can be accumulated in the voids of bentonite buffer around the disposal canister if gas generation rates become larger than the gas diffusion rate of bentonite buffer with the low-permeability. Continuous gas accumulations result in the increase in gas pressure, causing sudden dilation flow of gases with the gas pressure exceeding the gas breakthrough pressure. Given that the gas dilation flow can cause radionuclide leakage out of the engineered barrier system, it is necessary to consider possible damages affected by the radionuclide leakage and to properly understand the complicated behaviors of gas flow in the bentonite buffer with low permeability. In this study, the coupled hydro-mechanical model combined with the damage model that considers two-phase fluid flow and changes in hydraulic properties affected by mechanical deformations is applied to numerical simulations of 1-D gas injection test on saturated bentonite samples (refer to DECOVALEX-2019 Task A Stage 1A). To simulate the mechanical behavior of microcracks which occur due to the dilation flow caused by increase in gas pressure, a concept of elastic damage constitutive law is considered in the coupled hydro-mechanical model. When the TOUGH-FLAC coupling-based model proposed in this study is applied, changes in hydraulic properties affected by mechanical deformations combined with the mechanical damage are appropriately considered, and changes in gas injection pressure, pore pressures at radial filters and outlet, and stress recorded during the gas injection test are accurately simulated.

The backfill close the deep geological disposal system by filling the disposal tunnel and the connecting tunnel after the installation of buffer in the disposal hole. SKB and Posiva have established and designed the safety function of the backfill for the common goal of the deep geological disposal system. The safety function of backfill material has been set hydraulic conductivity of less than 10−10 m·s−1, a swelling pressure of 0.2 MPa, a compressive modulus of 10 MPa or a buffer density of 1,950 kg·m−3 or more, and freezing resistance. For the selection of the optimum backfill material, SKB and Posiva developed the concept of the backfill and evaluated the candidate that satisfies the requirements in four steps. In the first step, the performance and function that the backfill material should have were conceptualized. For the second step, laboratory tests and in-depth analysis of the candidate material properties were conducted. At this step, the focus has been on testing with the concept of the block method, using key candidate materials. In step 3, laboratory and large-scale experiments were performed to test engineering feasibility. In addition, design specifications for backfill materials were set based on site conditions, installation methods, and short- and long-term functions of materials. In Korea, it is only now in the step of selecting the concepts of the safety function. Therefore, it is necessary to benchmark the development process based on the previous studies of SKB and Posiva. In this study, candidate materials, experimental methods, and results were analyzed. As a result, the research steps and conditions for the selection of the optimum backfill material were reviewed. Using this study, the research steps of domestic backfill was suggested to develop within a short time for the Korean deep geological disposal system.

The timescale for the post-closure safety assessment of a deep geological repository ranges from ten thousand to a million year. In such a long period of time, the biosphere inevitably undergoes changes. Therefore, the long-term evolution of a biosphere is recognized as an important issue in the post-closure safety assessment of a deep geological repository for spent fuels. In this study, we reviewed the approaches to address the long-term evolution of a biosphere. The major drivers of longterm evolution of a biosphere are the climate change and the resulting landscape development. They can affect the hydrogeological and hydrogeochemical characteristics of a biosphere, and then the radionuclide migration through the biosphere followed by the exposure doses for the critical groups. In addition, human activities and the social developments can affect the climate change resulting in the long-term evolution of a biosphere. To make a biosphere assessment, the long-term evolution scenarios for the biosphere should be formulated considering these climate change, landscape development, and human activities. In addition, features, events, and processes (FEPs) that affect the long-term evolution of a biosphere should be used. According to the Safety Case reports of Finland, the major long-term evolution scenario drivers of a biosphere are local sea-level change due to climate change and land use related to crop type, irrigation procedures, livestock, forest management, construction of a well, and demographics. The climate change causing the local sea-level change can be simulated using various earth system models such as CLIMBER-2, MPI/UW, and UVic and an icesheet model such as SICOPOLIS. The review results of this study and FEPs related to the climate change, the landscape development, and human activities will be used to formulate long-term evolution scenarios for the safety assessment of a deep geological repository for spent fuels.

Geologic disposal of high-level radioactive waste is considered the most effective method to isolate high-level radioactive waste from the biosphere. A high-level radioactive waste repository is designed to be placed at a deep depth and generally consists of canisters, buffer material, and host rock. In the disposal system, the heat from the canister occurs for millions of years due to the long half-life of the high-level radioactive waste, and the heat induces vaporization of groundwater in the buffer material. The resaturation process also occurs due to groundwater inflow from the host rock by the hydraulic head and capillarity. The saturation variation leads to the heat transfer and multi-phase flow in the buffer material, and thermal pressurization of groundwater due to the heat affects the effective stress change in the host rock. The stress change can make the porosity and permeability change in the flow system of the host rock, and the flow system affects the nuclide migration to the biosphere. Therefore, it is crucial to understand the complex thermo-hydro-mechanical-chemical (THMC) coupled behavior to secure the repository’s long-term safety. DECOVALEX is an international cooperating project to develop numerical methods and models for predicting the THMC interactions in the disposal systems through validation and comparison with test results. In Task C of DECOVALEX-2023, nine participating groups (BGR, BGE, CAS, ENSI, GRS, KAERI, LBNL, NWMO, Sandia) models the full-scale emplacement (FE) experiments at the Mont Terri underground rock laboratory and focus on understanding pore pressure development, heat transfer, thermal pressurization, vaporization and resaturation process in the disposal system. In the FE experiment, three heaters generated heat with constant power for five years at a 1:1 scale in the emplacement tunnel based on Nagra’s reference repository design. KAERI used OGS-FLAC3D for the numerical simulation, combining OpenGeoSys for TH simulation and FLAC3D for M simulation. We generated a full-scale three-dimensional numerical model with a dimension of 100 by 100 by 60 meters. The pressure and temperature distribution were well simulated with the host rock's anisotropy. Based on the capillarity, we observed vaporization and resaturation in the bentonite under the twophase flow system. We plan to compare the simulation results with the field data and investigate the effect of input parameters, including thermal conductivity and pore compressibility affecting the thermal and flow system.

Deep geologic disposal of high-level nuclear wastes (HLW) requires intensive monitoring instrumentations to ensure long-term security. Acoustic emission (AE) method is considered as an effective method to monitor the mechanical degradation of natural rock and man-made concrete structures. The objectives of this study are (a) to identify the AE characteristics emitted from concretes as concrete materials under different types of loading, (b) to suggest AE parametric criteria to determine loading types and estimate the failure stage, and finally (c) to examine the feasibility of using AE method for real-time monitoring of geologic disposal system of HLW. This study performs a series of the mechanical experiments on concrete samples simultaneously with AE monitoring, including the uniaxial compression test (UCT), Brazilian tensile test (BTT) and punch through shear test (PTST). These mechanical tests are chosen to explore the effect of loading types on the resulting AE characteristics. This study selects important AE parameters which includes the AE count, average frequency (AF) and RA value in the time domain, and the peak frequency (PF) and centroid frequency in the frequency domain. The result reveals that the cumulative AE counts, the maximum RA value and the moving average PF show their potentials as indicators to damage progress for a certain loading type. The observed trends in the cumulative AE counts and the maximum RA value show three unique stages with an increase in applied stress: the steady state stage (or crack initiation stage; < 70% of yield stress), the transition stage (or damage progression stage; 70–90% of yield stress) and the rising stage (or failure stage; > 90% of yield stress). In addition, the moving average PF of PTST in the early damage stage appears to be particularly lower than that of UCT and BTT. The loading in BTT renders distinctive responses in the slope of the maximum RA–cumulative AE count (or tan ). The slope value shows less than 0.25 when the stress is close to 30% of BTT, 60% of UCT and 75% of PTST and mostly after 90% of yield stress, the slope mostly decreases than 0.25 in all tests. This study advances our understanding on AE responses of concrete materials with well-controlled laboratoryscale experimental AE data, and provides insights into further development of AE-base real-time diagnostic monitoring of structures made of rocks and concretes.