The treatment of solid radioactive waste can be divided into Mechanical (compaction), Thermal (Plasma), Melting (metal), Chemical (e.g. acid digestion) and Biochemical (e.g. bacteria). Among them, industrial thermal technologies include geomelt, Vitrificaion, Hip Ceramic, Incinerator, Pyrolysis, Plasma and Melting. In this study, the characteristics, status and advantages of geomelt vitrification were reviewed. Vitrification has long been considered an ideal choice for high-level radioactive waste by regulators internationally, because of its expected durability over hundreds of thousands of years. Geomelt vitrification is a highly flexible technology for hazardous and radioactive waste treatment. Uses electricity to melt waste materials to either destroy or immobilize contaminants. Final product is identical to natural obsidian very durable and resistant to weathering Geomelt vitrification creates ultra stable glass that is typically 10 times stronger than concrete, and more durable than granite or marble. Its leach resistance is among the highest of all materials in the world. In addition, contaminated soil, sludge, metals, organic matter, and bulky D&D debris can be treated simultaneously without pretreatment steps such as size reduction and sorting. Geomelt vitrification can be deployed in variety of in ground, in container or hybrid in cell treatment. Geomelt vitrification have been treating radioactive waste and hazardous waste since the 1990s, treatment in the U.S., UK, Australia, Japan and other countries. Initially developed by Pacific Northwest National Laboratory in the U.S., GeoMelt vitrification has been used successfully around the world for the U.S. Department of Energy (DOE) in Hanford and at Sellafield in the UK.

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.

With the rapid growth of nuclear power in China, a large number of dry wastes, which mainly include the high efficiency particulate air filters (glass fiber), cotton, polyethylene, and absorbent paper with low-level radioactivity and high volume, will be produced during the operation and maintenance of the nuclear power plants. Thermal plasma treatment is a world acceptable technology to incinerate and immobilize radioactive wastes, owing to the high volume reduction factor and the excellent chemical durability of the vitrified waste form. China has developed thermal plasma technology for the treatment of dry wastes from nuclear power plants for more than 15 years and the pilot plant has been constructed. This work will concentrate on the formulation of waste glass fiber to adapt to the vitrification process. A three-component (glass fiber-CaO-Na2O) constrained-region mixture experiment was designed and their viscosity data was mainly studied. The quadratic Scheffé model was used to plot the component effect on melting temperature. The retentions of simulated nuclides, such as Co, Sr, and Cs in the glasses were analyzed. In addition, the glass fiber as a glass matrix to immobilize residual ashes from the thermal plasma gasification of cotton, polyethylene, and absorbent paper was investigated as well.

During nuclear waste vitrification, loss of sodium (Na) and boron (B) occurs, as these elements are highly volatile at high temperatures, which causes fluctuations in composition and consequently affects the properties of the glass products. In this study, we investigated the volatilization behaviors of Na and B from a simulated high-level waste glass as functions of heating temperature and dwelling duration. Based on the data obtained regarding the composition of Na and B and the structure of the glass, a hypothetical model was proposed to explain the volatilization behaviors of Na and B from a structural viewpoint. As the loss of Na and B during vitrification, the crystallization of the glass occurred. Thus, the crystallization behavior of the simulated waste glass upon composition deviation was studied.

Hanford site has been operated since 1943 to produce the plutonium for nuclear weapons. Significant amount of radioactive wastes was generated by the nuclear weapons production process. The radioactive wastes are stored in 177 aged underground tanks. Due to the risk of leakage into the air and the Columbia River, the US DOE and EPA, and Washington State Department of Ecology organized the Tri-Party Agreement (TPA) to clean-up the Hanford site in 1989. The LAW (low-activity waste) vitrification facility named WTP (Waste Treatment Plant) is plan to vitrify about 212 million liters of radioactive waste. The US DOE announced that the world’s largest melter to vitrify the LAW was heated up on October 8, 2022.

KHNP’s vitrification technology introduced a commercialized vitrification facility to the Hanul nuclear power site after a commercialization test through a lab test and a pilot plant at KHNP-CRI. France’s ANADEC (consortium with CEA, Orano, ECM Technologies and Andra) conducted a feasibility evaluation from FY2018 to FY2021 to apply In-Can vitrification, which was developed to treat Fukushima Effluent Treatment Waste (FETW) such as carbonate slurry and ferric slurry generated from ALPS (Advanced Liquid Processing System-Multi Radionuclides Removal) facilities for waste treatment in Fukushima, Japan. For commercialization, the following method was used. First, through the Laboratory scale studies, the possibility of high waste loading (60wt% in dry mass) of slurry on borosilicate matrix was tested. In addition, the volatility of radionuclide was evaluated through radionuclides surrogates with a Bench-scale mockup and glass discharge (100 kg) was evaluated through In-Can vitrification process verification. The feeding system was improved through a pilot scale test, and finally, glass discharge (300 kg) was evaluated after large amount of waste was treated through an industrial prototype (Fullscale) at the CEA Marcoule site (France).

A vitrification facility control area is formed to control and monitor the vitrification facility process, and the control system is designed to manage the vitrification facility more safely and effectively. The control system is largely composed of a process control system and an off-gas monitoring system. The process control system is operated so that operation variables can be maintained in a normal state even in normal and transient conditions, and is designed so that the vitrification facility can be stably maintained in the event of an abnormality in the facility. The process control system consists of Programmable Logic Controller (PLC) and Local Control Panel (LCP), which controls and monitors each unit device. In addition, operation variables are provided to the operator so that the operator can manage operation variables during process control in a centralized manner for the operation of the vitrification facility. The off-gas monitoring system is operated to monitor whether the off-gas discharged to the environment is stably maintained within the standard level, and the off-gas is monitored through an independent monitoring system.

After melting glass at a high temperature of about 1,100 degrees in the Cold Crucible Induction Melter (CCIM) of the vitrification facility, radioactive waste is fed into the CCIM to vitrify radioactive waste. Accordingly, since the metal sector of the CCIM contacts the high-temperature molten glass, cooling water is supplied to continuously cool the metal sector. The cooling system is divided into primary and secondary cooling water systems. The primary cooling water flows inside the metal sector of the CCIM to maintain the metal sector within normal temperature, thereby forming a glass layer between the metal sector and the high-temperature melting glass. The secondary cooling system is a system that cools the primary cooling water that cools the metal sector, and removes heat generated from the primary cooling system. In addition, it is designed to stably supply cooling water to the secondary cooling water system through an emergency cooling water system so that cooling water can be stably supplied to the secondary cooling water system in the event of secondary cooling water loss. Therefore, it is designed to maintain the facility stably in the event of loss of cooling water for the CCIM of the vitrification facility.

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.

Radioactive waste generated in various forms needs to be technically stabilized for safe treatment, disposal, and long-term safety. Overseas, research on vitrification with excellent mechanical and physicochemical properties of high-level waste is being actively conducted. Vitrification is a process of converting radioactive waste into a stable and durable glass-like material, which is then safe storage and disposal. The stability of vitrified form is an important concern for environment safety, as any leakage or release of radionuclide could have serious consequences for health and the environment. Therefore, several studies are being conducted on the disposal stability of vitrification of radioactive waste. In order to evaluate the stability of the solidified form, mechanical properties such as density, microhardness, and compressive strength and chemical durability such as leaching properties should be performed. There are several types of leaching test methods to evaluate the chemical durability, which is important in the characterization of solidified forms. In this study, the leaching test method for chemical durability evaluation and the evaluation results of leaching characteristics according to pH were reviewed.

In operating or permanently shut down nuclear power plants which were built between 1970s and 1990s, asbestos was widely used for ceiling materials, wall materials, and gaskets. Furthermore, it was mainly treated as a heat-resistant material like insulation. In Kori Unit 1, radioactive asbestos was replaced or removed through maintenance and repair in the containment building during the operation period of about 40 years, but radioactive asbestos still remains that need to be partially dismantled. Generally, it is more difficult to handle because it belongs to two different waste categories, radioactive waste and hazardous waste. In addition, the risk increases further due to radioactivity with the asbestos hazards itself. Therefore, it is very important to accurately determine the amount of radioactive asbestos waste and to evaluate the treatment method and disposal reduction rate before the decommissioning is started. According to the Korean Waste Management Act, three methods are recommended for the asbestos (hazardous waste) treatment: landfill, solidification, and high-temperature melting. Landfill is commonly used in Korea and the United States while high-temperature melting and solidification are additionally recommended only in Korea. Considering the situation in Korea, landfill is not appropriate due to the limitations of landfill capacity and potential risks (hazards still remain). Therefore, the other two methods can be considered sufficiently in terms of safety, detoxification, and reduction rate. This paper evaluates the amount of radioactive asbestos waste at Kori Unit 1 based on the actual asbestos building material data (as of February 2022) of the Asbestos Management Comprehensive Information Network. Vitrification is considered as a sufficient alternative for treating radioactive asbestos waste. And, it is checked whether the vitrified waste through the high-temperature melting method, plasma torch, meets the requirements such as detoxification, compressive strength and leachability for storage and disposal stability. It is expected to be useful to prepare a radioactive mixed waste management standard and to reduce the disposal cost through the reduction of final waste.

Spent nuclear fuels are temporarily stored in nuclear power plant site. When a problem such as cracking of spent nuclear fuel assembly or cladding occurs or uranium that has not been separated during the reprocessing remains, it is necessary to treat it. The borosilicate glasses have been considered to vitrify whole spent nuclear fuel assembly. However, a large amount of Pb addition was necessary to oxidize metals in assembly to make them suitable for oxide glass vitrifcation. Furthermore, these borosilicate glasses need to be melted at high temperatures (> 1,400°C) when UO2 content is more than 20wt%. Iron phosphate glasses can be melted at a relatively low temperature (< 1,300°C) even with a similar UO2 addition. A composition of iron phosphate glass for immobilization of uranium oxide has been developed. The glasses have glass transition temperatures of ~555°C that are high enough to maintain its phase stability in geological repositories. The waste loading of UO2 in the glass is ~33.73wt%. Normalized elemental releases from the product consistency test were well below the US regulation of 2 g/m2. Nuclear criticality safety and heat generation in deep geological repositories were calculated using MCNP and computational fluid dynamics simulation, respectively. The glass had effective neutron multiplication factor (keff) of 0.755, which is smaller than the nuclear- criticality safety regulation of 0.95. Surface temperature of the disposal canister is expected to lower than the limit temperature (< 100°C). Most of the U in the glass is in the 4+state, which is more chemically durable than the 6+state. As a result of long-term dissolution experiment, chemically-durable uranium pyrophosphate (UP2O7) crystals were formed.

Cryopreservation of porcine ovarian tissue by vitrification method is a promising approach to preserve genetic materials for future use. However, information is not enough and technology still remains in a challenge stage in pig. Therefore, the objective of present study was to determine possibility of vitrification method to cryopreserve porcine ovarian tissue and to confirm an occurrence of cryoinjuries. Briefly, cryoinjuries and apoptosis patterns in vitrified-warmed ovarian tissue were examined by histological evaluation and TUNEL assay respectively. In results, a damaged morphology of oocytes was detected among groups and the rate was significantly (p < 0.05) lower in vitrification group (25.8%) than freezing control group (67.7%), while fresh control group (6.6%) showed significantly (p < 0.05) lower than both groups. In addition, cryoinjury that form a wave pattern of tissues around follicles was found in the frozen control group, but not in the fresh control group as well as in the vitrification group. Apoptotic cells in follicle was observed only in freezing control group while no apoptotic cell was found in both fresh control and vitrification. Similarly, apoptotic patterns of tissues not in follicle were comparable between fresh control and vitrification groups while freezing control group showed increased tendency. Conclusively, it was confirmed that vitrification method has a prevention effect against cryoinjury and this method could be an alternative approach for cryopreservation of genetic material in pigs. Further study is needed to examine the viability of oocytes derived from vitrified-warmed ovarian tissue.

The optimum vitrification conditions of the radioactive waste using high-temperature furnace and HIP (Hot Isostatic Press) were studied for the successful reduction of the solidification volume, radioactive level, satisfying the disposal criteria such as leaching rate and compressive strength. Vitrification is receiving attention for the solidification disposal of intermediate and low-level radioactive wastes for its chemical-physical stability and leachability. Its principle is to trap the radioactive material in a fixed structure of the glass type materials, such as Boron Trioxide, Silicon Dioxide and Phosphorus Pentoxide. Sludge targeted in this study is assembly of materials while sludge is stored in the stainless-steel tank before disposal, which consists of Fe3O4 (14.9wt%), Fe2O3 (3.8wt%), and Cr2O3 (6.3wt%), cement paste (25wt%) and detergent/shower sludge (50wt%). The detergent/shower sludge generated from the washing the clothes that were worn during the work at the laboratory and nuclear power plant contains organic materials that are vulnerable to chemical reactions, therefore, immobilization of organic material by the incinerating step, which can also immobilize the radioactive substance, was applied. Its composition – containing Cs-133 and Co-59 substitution for Cs-134 and Co-60 that are radioactive – was analyzed by XRD before and after the mineralization of the sludge using high temperature furnace in different temperature, to identify the remaining element and the features of the mineralized sludge. Targeted sludge was vitrificated using Hot Isostatic Press in with different pressure and temperature conditions, to find out the optimum vitrification conditions. Vitrificated waste was evaluated in many aspects - leaching evaluation following ANS16.1, compressive strength evaluation of 3.44 MPa (waste disposal criteria), volume reduction before and after the sequence.