Safety for the radioactive waste disposed of in the disposal facility should be secured through safety assessment in consideration of the various situations. In this study, the influence and correlation of EDTA and ISA, which are the factors that can impede the safety of the disposal facility, were analyzed using the PHREEQC computational code. Thermodynamic database (TDB) of Andra, specific ion interaction theory (SIT) model as ionic strength correction model, radionuclides (Ni, Am, Pu) were adopted to perform the calculation on the distribution of chemical species by pH. According to the results, EDTA dominated the system and the effect of ISA is relatively small for the distribution of the chemical species of divalent and trivalent cations in neutral and weak base conditions. In the case of the tetravalent cations, the effect of ISA increased compared to the previous case especially in the strong base conditions. In conclusion, EDTA has a more significant effect on the system than ISA under the environment of the domestic disposal facility. Furthermore, when EDTA and ISA are present simultaneously in the system, the effects of two materials are inversely proportional and this characteristic should be considered during the safety assessment.

In a recent preliminary inspection for disposal, the glass fiber waste (GFW), used as a pipe insulation, was judged as “pending evaluation” because some dust was found in drum opening tests. Therefore, additional inspection is required to ensure that the package corresponds with the acceptance criteria of the particulates. The dust was generated presumably due to GFW being used in a high-temperature environment for a long time, thus being easily degraded and crushed. For this reason, safety issues that may occur in the process of handling, transportation, and disposal are emerging. Therefore, in this study, a preliminary safety assessment of GFW disposal was performed, the exposure dose to the general public was derived, and compared with the dose limit. The evaluation was carried out in the following order: (1) evaluation of GFW radiation source term, (2) selection of accident scenario, (3) calculation of exposure dose, (4) comparison of evaluation results with dose limits, and confirmation of satisfaction. The average radioactivity of the GFW to be disposed of was used as the source term, and the main nuclides were identified as H-3, Fe-55, Co-60, Ni-63, and Pu-241. In general, the types of accidents that can occur at disposal facilities can be classified into falls, fires, collisions during transportation, off-site accidents, and nuclear criticality, and the accident scenarios are selected by analyzing and reviewing the probability of each accident. In this study, the accident analysis and scenarios presented in the safety assessment of the KORAD were reviewed, and the fire in the treatment facility, the fire in the storage facility, and the collision of the transport vehicle were selected as the evaluation scenarios. When an accident occurs, the radioactive material inside the container leaks out and diffuses into the atmosphere. In this evaluation, the internal and external exposure of the general public due to radioactive plume at the site boundary was evaluated and the dose conversion factors from ICRP-72 and FGR 12 were used. Based on the evaluation, general public was exposed to 0.004 mSv, 0.013 mSv, and 0.045 mSv, respectively, due to a fire at a treatment facility, at a storage facility, and in a transport vehicle. Most of the dose is due to internal exposure by Pu-241 nuclide, because the proportion of it in the waste is high, and when inhaled, the internal dose is high by emitting beta rays. It was confirmed that the result of dose was 0.4%, 1.3% and 4.5% of the annual dose limit, sufficiently satisfying the dose limit and safety.

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

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Thus, evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, safety assessments conducted in Sweden and UK are studied, and the factors required to be considered for appropriate safety assessment of cellulose is analyzed. SKB (Sweden) conducted safety assessment of cellulose degradation as a part of long-term safety assessment of SFR. SKB determined that ISA would impact sorption of trivalent and tetravalent radionuclides (Eu, Am, Th, Np, Pa, Pu, U, Tc, Zr and Nb) at concentration higher than 10−4–10−3 M, and impact sorption of divalent radionuclides (Ni, Co, Fe, Be and Pb) at concentration higher than 10−2 M. Then, SKB conservatively set the upper limit of ISA concentration to be 10−4 M and conducted cellulose degradation evaluation on each waste package type, considering the expected disposal environment of SFR. Based on the calculated results, some of the waste packages showed concentration of ISA to be higher than 10−4 M, so SKB conservatively developed waste acceptance criteria to prevent ISA being produced to an extent of affecting the safety of the repository. SKB conducted safety assessment only for the repositories with pH above 12.5 and excluded 1BLA from the safety assessment as the expected pH of 1BLA is around 12, which is insufficient for cellulose to degrade. However, SKB set disposal limit for 1BLA as well, to minimize potential impact in future. Serco (UK) conducted safety assessment of cellulose degradation for the conceptual repository, which is a concrete vault with cementitious backfill. Serco estimated that the pH of repository would maintain around 12.4. Serco conservatively assumed that the pH would be sufficient for cellulose degradation to occur partially, and suggested application of appropriate degradation ratio for safety assessment of cellulose degradation. To conduct appropriate safety assessment of cellulose degradation, an appropriate ISA concentration limit based on radionuclide inventory list, and an appropriate cellulose degradation ratio based on the pH of disposal environment should be determined. As for guidance, below pH 12.5, cellulose degradation is not expected, and between pH 12.5–13, partial cellulose degradation is expected. In future, this study could be used as fundamental data to evaluate safety of the repository.

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

In this introduction, test devices for radwaste characterization specimen was developed and utilized. In order to permanently dispose of solidified radwastes, not only radioactive characterization but also physical & chemical characterization shall be performed to assess compliance with the waste acceptance criteria. Waste acceptance criteria can be made up measurement of free standing water, compressive strength test, thermal cycling test, radiation resistance test, leaching test, immersion test. Approximately, the equipment for each test is sorted out five types. equipment for making characterization specimen, equipment for compressive strength test, equipment for thermal cycling test, equipment for radiation resistance test, equipment for Immersion test and leaching test. Equipment for making characterization specimen is operated the dry process. The equipment of two types: one (sampling device) that cores solidified radioactive waste in a drum, and the other (cutting machine) that properly cuts the coring samples. Sampling device is not used in industry, so it is specially manufactured, cutting machine is using ready-made products. In addition, devices for compressive strength test and thermal cycling test are using ready- made products. Facility for Radiation resistance test is located in Jeong-eup. For the efficient test, a table was manufactured in the columnar form like the specimen. Finally, devices for immersion test and leaching test are so transformed that contact all surfaces of the specimen with the liquid.

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.

Radioactive waste disposal facility in Korea, radioactive waste packaged in 200 L drums is placed in a concrete disposal container and disposed of at an underground silo type (cave) disposal facility. At this time, the disposal container cover is seated on the top of the disposal container, and if the disposal container and the cover are not completely combined, the container cover is raised up from the top of the disposal container, so safety problems may occur when stacking the disposal container. Therefore, various methods exist to secure a margin for the pure height inside the disposal container. The disposal container cover only covers the upper surface of the container to shield radiation, and structural performance is not required. Therefore, the method of processing the cover, such as a method of making the cover of the disposal container thin, is the easiest method to apply. In this study, a method to reduce the thickness of the cover of a concrete disposal container was devised, and structural performance under usability conditions such as lifting and seating was analyzed. In addition, the disposal container cover has a reinforced concrete form in which dissimilar materials (concrete and steel) are combined, an integrated analysis was performed to secure the reliability of the analysis results for this, and the analysis results were described. It was found that the proposed disposal container cover structure can improve usability by reducing the stress concentration phenomenon.

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

When the decommissioning of a nuclear power plant begins in earnest, starting with Kori Unit 1, it is necessary to dispose of intermediate-level wastes such as high-dose waste filters and waste resin stored in the power plant, as well as the internal structures of the reactor. However, there are no intermediate-level waste disposal facilities in Korea, and the maintenance of acceptance criteria considering the physical, chemical, and radiological characteristics of intermediate-level waste is insufficient. In this paper, in preparation for the establishment of domestic intermediate-level waste treatment/disposal and acceptance standards, the following major foreign countries’ legal and institutional standards for intermediate-level waste are reviewed, and based on this, factors to be considered when establishing domestic intermediate-level waste treatment/disposal standards were derived. First, although the USA does not define and manage intermediate-level wastes separately, low-level wastes were separated into Class A, B, and C, where land disposal is allowed, and GTCC, which does not allow land disposal. However, it was recently confirmed that the position was changed to recognize the possibility of land disposal of GTCC waste under the condition that the dose to inadvertent intruders does not exceed 5 mSv·yr−1 and a barrier against inadvertent intrusion valid for 500 years is installed. Second, Sweden classifies intermediate-level wastes into short-lived and longlived intermediate-level wastes. The maximum dose rate permitted on packages are different for each vault and a silo of the SFR where short-lived wastes; 100 mSv·h−1 or less is disposed of in BMA, 10 mSV·h−1 or less in BTF, 2 mSv·h−1 or less in BLA and 500 mSv·h−1 or less in silo. Meanwhile, a repository for long-lived low and intermediate level waste, SFL, which could contains significant amounts of nuclides with a half-life greater than 31 years, operations are planned to commence in 2045. Third, France also manages short-lived intermediate-level wastes and long-lived intermediatelevel wastes separately, and the short-lived intermediate-level wastes were disposed of together with short-lived low-level wastes at the La Manche and L’Aube repository. France announced the Cigéo Project, a high- and medium-level long-lived waste plan in 2012, and submitted the creation authorization application for in 2021 with the goal of operating a repository in 2025. Finally, the UK defines intermediate-level waste as “waste whose activity level exceeds the upper limit for low-level waste but does not require heating, which is considered in the design of storage or disposal facilities” and established NIREX to provide deep disposal of intermediate-level radioactive waste. In Finland, wastes with radioactive concentrations of 1 MBq/kg to 10 GBq·kg−1 are classified as intermediatelevel wastes, and a repository was constructed and operated in a bedrock of about 110 m underground. Because the domestic classification standard simply classifies intermediate-level waste as waste exceeding the activity level of low-level waste limit, not high-level wastes, it is necessary to establish treatment and disposal standards by subdividing them by dose rate and long-lived radionuclides concentration to safely and efficiently dispose of intermediate-level waste for. Additionally, there is a need to decide whether or not to reflect safety by inadvertent intruders when evaluating the safety of intermediate-level disposal.

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

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

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, the maximum safe disposal amount cellulose is evaluated considering the disposal environment of silos of 1st phase disposal facility. The key factor governing the impact of cellulose degradation is pH of disposal environment, as cellulose is known to degrade partially at pH above 12.5, and completely at pH above 13. Thus, disposal environment should be analyzed as to determine the extent of degradation. As silos are constructed with large amount of cement, porewater within concrete walls would be of very high pH. However, for high pH porewater to be released into the pores of crushed rock, which is filling up the silos, lower pH groundwater (commonly pH 7) should flow into the silos through the concrete walls. This causes dilution of the high pH concrete porewater, resulting in a lower pH as the silos are filled, reaching to expected pH of 11.8–12.3, which is below cellulose degradation condition. Thus, cellulose degradation is not expected, but to quantitatively evaluate safe disposal amount of cellulose, partial degradation is assumed. Upon literature review, the most conservative ISA concentration, enhancing radionuclide mobility, is determined to be 1.0×10−4 M and to reach this concentration, cellulose mass equivalent to 6wt% of cement of the repository, is required to be degraded. However, this ratio is derived based on complete degradation of cellulose into ISA, so for partial degradation, degradation ratio and yield ratio of ISA should be considered. Commonly, cellulosic material (e.g. cotton, paper, etc.) has degree of polymerization (DP) between 1,000–2,000, and with this DP, degradation ratio is estimated to be about 10%. Furthermore, yield ratio of ISA is known to be 80%. Considering all these aspects, about 1.79×107 kg of cellulose could be disposed, which if converted into number of drums, considering cellulose content of dry active waste, more than 100,000 drums (200 L) could be disposed with negligible impact on safety. Based on the result, negligible impact of cellulose degradation is expected for safety of 1st phase disposal facility. In future, this study could be used as fundamental data for revising waste acceptance criteria.

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

Mechanism and kinetics of Rhenium complexes as a surrogate of Technetium-99 (99Tc) is worthy of study from radioactive waste safe disposal perspective. Re(IV)-EDTA was synthesized via the reduction of Re(VII) with Sn(II) in the presence of Ethylenediaminetetracetic acid (EDTA). The Re(IV)-EDTA was then degraded by H2O2 (7–30%) at pH of 3–11 in ionic strength I = 0–2 M solution. The Re- EDTA was observed to degrade more rapidly at pH of ≤ 3–4 than one of ≥ 10–11 and remained stable at pH = 7–9. At a low acidic pH, the complex degradation process was facilitated by protonation and corresponded to the exponential model (y = k. e–nt). In contrast, at an alkaline pH, the degradation was facilitated OH– complexation with Re(IV) and corresponded to a linear model (y = –mt + C). Complex degradation followed the zero-order rate kinetics for the H+ and Re-EDTA parameters, apart from a pH of 3, for which degradation was a better fit to first order kinetics. A higher Re(IV)-EDTA stability at a pH of 7–9 demonstrated that Re(IV)-EDTA (or 99Tc(IV)-EDTA) tends to be more persistent in natural environmental conditions.

Expansive clays (for examples, bentonites) are favored as buffer and backfill materials because of their low hydraulic conductivity, high swelling potential, and good mechanical properties, and are installed in highly compacted blocks in repositories. Compacted expansive clays have a dual-structure system: macrostructural system which is a complex of clay aggregates with the inter-aggregate pores (macropores) which can be filled by either liquids or gases; microstructural system with the intraaggregate pores between or within clay particles (micropores) which is usually considered to be saturated by liquid. Understanding the dual-strucure system of expansive clays is essential for characterizing and modeling multiphysics (stress-strain, swelling pressure, etc.) in buffers and backfills. Existing multiphysics studies of expansive clays, as in non-expansive soils, were mostly conducted with a single structure approach based on the behavior of macropores, and there have been limitations in the comprehensive interpretation and modeling of experimental results. However, with the recent development of measurement techniques, a lot of available information on the pore structure of compacted expansive clays has been reported, and with the results, a dual-structure approach considering both microstructural and macrostructural systems has been increasingly applied to improve the modeling of multiphysics of expansive clays. This study reviewed the dual-structure system of compacted expansive clays, analyzed previous studies on its evolution according to hydromechanical loading (loading-unloading and wetting-drying paths), and based on these, intended to provide technical knowledge and information needed for multiphysics research of expansive clays-based buffer and backfill for the KRS repository.

In order to monitor the long-term condition of structures in nuclear waste disposal system and evaluate the degree of damage, it is necessary to secure quantitative monitoring, diagnosis, and prediction technology. However, at present, only simple monitoring or deterioration evaluation of the structure is being performed. Recently, there is a trend to develop monitoring systems using artificial intelligence algorithms, such as to introduce artificial intelligence-based failure diagnosis technology in nuclear power plant facilities. An artificial intelligence algorithm was applied to distinguish the noise signal and the destructive signal collected in the field. This can minimize false alarms in the monitoring system. However, it is difficult to apply artificial intelligence to industrial sites only by learning through laboratory data. Therefore, a database of noise signals and destructive signals was constructed through laboratory data, and signals effective for quantitative soundness determination of structures were separated and learned. In addition, an adaptive artificial intelligence algorithm was developed to enable additional learning and adaptive learning using field data, and its performance was verified through experiments.

The geological disposal of spent nuclear fuel is one of the important problems to be solved worldwide. For the safety of the geological disposal, disposal facility is recommended to be constructed in the deep reducing environment of host rocks. As host rocks, rock salt, argillaceous (clay) rock, and crystalline rock have been considered as stable geological formations in various countries. Although various studies have been conducted on crystalline rocks in Korea, there are still few studies on hydrogeochemical evolution in the deep and reducing environment related to the disposal of spent nuclear fuel. Therefore, this study was conducted to identify hydrogeochemical evolution process in granite aquifer which can affect the stability of disposal facility. Groundwater samples for isotope and chemical analysis were collected quarterly adjacent to KURT (KAERI Underground Research Tunnel). As the depth increased, the groundwater changed from Ca-HCO3 type to Na-HCO3 type under the influence of silicate mineral weathering, and the fluorine concentration increased due to the dissolution of fluorine-bearing minerals. However, hydrogeochemical evolution according to the depth was not observed in some wells because of a hydraulic connection through the fracture zone. In addition, the behavior of nitrate and redox-sensitive metals (Fe, Mn, U, Mo) in groundwater was clearly different in the redox condition. Considering these hydrogeochemical processes and hydrogeological factors, a conceptual model of granite aquifers in and around KURT was established. The results of this study will be used as basic data to understand the hydrogeochemical processes and to evaluate and predict the behavior of radionuclides in granite aquifer system.

High level nuclear waste (HLW) is surely disposed in repository in safe by being separated from human life zone. Deep geological disposal method is one of the most potent disposal method. Deep geological repository is exposed to high pressure and groundwater saturation due to its depth over 500 m. And it is also exposed to high temperature and radiation by spent fuels. Thus, HLW repository suffers extremely complex thermo-hydro-mechanical-radioactive condition. Long-term integrity of repository should be verified because the expected lifetime of the repository is over 10,000 years. However, the integrity of monitoring sensors are not reach the endurance lifetime of the repository with present technology. And the disposal condition, thermo-hydro-mechanical-radioactive, should shorten the estimated lifetime of the monitoring sensors. Therefore, it is necessary to improve the long-term integrity of the monitoring sensors. Although long-term tests are required to identify the prolonged durability of monitoring sensors, accelerated tests can help curtail test period. Accelerated tests is classified into accelerated stress test and accelerated degradation test and their methodology and theories are investigated. Their tests are design and proceed by following process: 1) identify failure modes, 2) select accelerated stress parameter, 3) Determine stress level, 4) Determine testing time and number of specimens, 5) Define measurement paremeter and failure criteria, 6) Suggest measurement method and measurement duration. Literature reviews were conducted to identify the influence of the disposal conditions such as thermo-hydro-mechnical-radioactive on integrity of material and monitoring sensors. The investigated data reported in this paper will be utilized to verify the improvement of integrity of monitoring sensors.

The criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of PWR SNF disposal canister: the KBS-3 type canister and the small standardized transportation, aging and disposal (STAD) canister. The criticality analyses were carried out for four cases as follows: (1) the calculation of isotopic compositions within a SNF using a depletion assessment code and (2) the calculation of the effective multiplication factor (keff) value using a criticality assessment code. Firstly, the KBS-3 type canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.74407, 0.74102, and 0.73783, respectively. Secondly, the STAD canister was modelled. The SNFs contained in the STAD canister were assumed to be the enrichment of 4.0wt% and the burnup of 45,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were estimated to be 0.71448, 0.70982, and 0.70743, respectively. Thirdly, the KBS-3 canister with four SNFs of which the enrichment was 4.5wt% and the burnup was 55,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.73366, 0.72880, and 0.72634, respectively. Finally, the calculations were carried out for the STAD canister containing four SNFs of the enrichment of 4.5wt% and the burnup of 55,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were 0.70323, 0.69946, and 0.69719, respectively. Therefore, all of four cases met the performance target with respect to the keff values, 0.95. The STAD canister showed lower keff values than the KBS-3 canister. This appears to be the neutron absorber plate installed in the STAD canister although the distance among the four SNFs in the STAD canister was shorter than the KBS-3 canister.