Various dry active wastes (DAWs) have been accumulated in nuclear power plants since the DAWs are mostly combustible. KAERI has developed a thermochemical treatment process for the used decontamination paper as an operational waste to substitute for incineration process and to decontaminate radionuclides from the DAWs. The thermochemical process is composed of thermal decomposition in a closed vessel, chlorination of carbonated DAWs, separation of soluble chlorides captured in water by hydroxide precipitation, and immobilization of the precipitate. This study examined the third and fourth steps in the process to immobilize Co-60 by fabricating a stable wasteform. Precipitation behaviors were investigated in the chloride solution by adding 10 M KOH. It was shown that the precipitates were composed of Mg(OH)2 and Al(OH)3. Then, the glass-ceramic wasteform for the precipitates were produced by adding additive mixtures in which silica and boron oxide were blended with various ratios. The wasteform was evaluated in terms of volume reduction ratio, bulk density, compressive strength, and leachability.

A study was conducted on the vitrification of the rare earth oxide waste generated from the PyroGreen process. The target rare earth waste consisted of eight elements: Nd, Ce, La, Pr, Sm, Y, Gd, and Eu. The waste loading of the rare earth waste in the developed borosilicate glass system was 20wt%. The fabricated glass, processed at 1,200℃, exhibited uniform and homogeneous surface without any crystallization and precipitation. The viscosity and electrical conductivity of the melted glass at 1,200℃ were 7.2 poise and 1.1 S·cm−1, respectively, that were suitable for the operation of the vitrification facility. The calculated leaching index of Cs, Co, and Sr were 10.4, 10.6, and 9.8, respectively. The evaluated Product Consistency Test (PCT) normalized release of the glass indicated that the glass satisfied the requirements for the disposal acceptance criteria. Furthermore, the pristine, 90 days water immersed, 30 thermal cycled, and 10 MGy gamma ray irradiated glasses exhibited good compressive strength. The results indicated that the fabricated glass containing rare earth waste from the PyroGreen process was acceptable for the disposal in the repository, in terms of chemical durability and mechanical strength.

Vitrification is one of the best ways to immobilize high-level radioactive waste (HLW) worldwide over the past 50 years. Since the glass matrix has a medium (3.0-5.5 A) and short (1.5-3.0 A) periodicity, it can accommodate most elements from the periodic table. Borosilicate glass is the most suitable glass matrix for vitrification due to its high chemical durability, high waste-loading capacity, and good radiation resistance. Mo is a fission product contained in liquid waste generated in the process of reprocessing spent nuclear fuel and exists in the form of MoO4 2- in the glass. MoO4 2- forms a depolymerization region without directly connecting with the glass network former. When the concentration of Mo increases in the depolymerization region, it combines with nearby alkali or alkaline earth cations to form a crystalline molybdate phase. Phase separation and crystallization in the glass can degrade the performance of the material because it changes the physical and chemical properties of the glass. In particular, since alkali molybdate has high water solubility when it forms crystals containing radioactive elements such as Cs, there is a risk of leakage of radionuclides by groundwater during deep underground disposal. Therefore, in this study, the most stable glass-ceramic composition was developed using various alkali elements, and the difference in phase separation and crystallization behavior in glass and the stability of the solidified body were analyzed by structural analysis of the glass network and alkali molybdate. The cause of the difference in crystallization of alkali molybdate according to the type of alkali cation is structurally analyzed, and using this, research is conducted to increase the Mo content in the glass without crystallization.

PURPOSES : The purpose of this study is to confirm the thermal expansion characteristics of concrete mixed with 1% waste glass fine aggregates, which is the amount stipulated for recycled aggregates in the current quality standard. METHODS : The coefficient of thermal expansion was measured by applying AASHTOT 336-10 using a LVDT. The results measured were used as physical properties in a finite element analysis to confirm the change in tensile stress and the displacement of the right angle section of the upper slab of a concrete pavement due to admixture substitution. RESULTS : The thermal expansion coefficients of concrete based on the replacement rate of the admixture when the waste glass fine aggregates are replaced are within the range of the thermal expansion coefficients of concrete specified in the Federal Highway Administration report. As the replacement rate of the admixture increases, the thermal expansion coefficient of concrete decreases. As the thermal expansion coefficient decreases, the slab pavement curling displacement and the tensile stress of the center of the upper slab of concrete decrease. CONCLUSIONS : In the short term, the presence or absence of waste glass fine aggregates does not significantly affect the thermal expansion coefficient of concrete. However, in the long term, waste glass fine aggregates are reactive aggregates that causes ASR, which creates an expandable gel around the aggregates and results in concrete expansion. Therefore, the relationship between ASR and the thermal expansion coefficient must be analyzed in future studies.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

Glass wool, the primary material of insulation, is composed of glass fibers and is used to insulate the temperature of steam generators and pipes in nuclear power plants. Glass fiber is widely adopted as a substitute for asbestos classified as a carcinogen. The insulations used in nuclear power plants are classified as radioactive waste and most of the insulation is Very Low-Level Waste (VLLW). It is packaged in a 200 L drum the same as a Dry Active Waste (DAW). In the case of the insulations, it is packaged in a vinyl bag and then charged into the drum for securing additional safety because of the fine particle size of the fiberglass. A safety assessment of the disposal facility should be considered to dispose of radioactive waste. As a result of analyzing overseas Waste Acceptance Criteria (WAC), there is no case that has a separate limitation for glass fiber. Also, in order to confirm that glass fibers can be treated in the same manner as DAW, research related to the diffusion of glass fibers into the environment was conducted in this paper. It was confirmed that the glass fiber was precipitated due to the low flow velocity of groundwater in the Gyeongju radioactive waste repository and did not spread to the surrounding environment due to the effect of the engineering barrier. Therefore, the glass fiber has no special issue and can be treated in the same way as a DAW. In addition, it can be disposed of in the disposal facility by securing sufficient radiological safety as VLLW.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

As the design life of nuclear power plants are coming to the end, starting with Kori unit 1, nuclear power related organizations have been actively conducted research on the treatment of nuclear power plant decommissioning waste. In this study, among various types of radioactive waste, stabilization and volume reduction experiments were conducted on radioactive contaminated soil waste. Korea has no experience in decommissioning nuclear power plants, but a large amount of radioactively contaminated soil waste was generated during the decommissioning of the KAERI research reactor (TRIGA Mark- II) and the uranium conversion facility. This case shows the possibility of generating radioactive soil waste from nuclear power plants and nuclear-related facilities sites. Soil waste should be solidified, because its fluidity and dispersibility wastes specified in the notification of the Korea Nuclear Safety and Security Commission. In addition, the solidified waste forms should have sufficient mechanical strength and water resistance. Numerous minerals in the soil are components that can make glass and ceramics, for this reason, glass-ceramic sintered body can be made by appropriate heat and pressure. The sintering conditions of soil were optimized, in order to make better economical and more stable sintered body, some additives (such as additives for glass were mixed) with the soil and sintering experiments were conducted. Uncontaminated natural soil was collected and used for the experiment after air drying. Moisture content, pH, bulk density, and organic content were measured to understand the basic properties of soil, and physicochemical properties of the soil were identified by XRD, XRF, TG, and SEM-EDS analysis. In order to understand the distribution by particle size of the soil, it was divided into Sand (0.05–2 mm) and Fines (< 0.05 mm). The green body was manufactured in the form of a cylinder with a diameter of 13mm and a height of about 10mm. Appropriate pressure (> 150 MPa) was applied to the soil to make a green body, and appropriate heat (> 800°C) was applied to the sintered body to make a sintered body. The sintering was conducted in a muffle furnace in air conditions. The volume reduction and compressive strength of the sintered body for each condition were evaluated.

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.

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.