Vitrification, one of the most promising solidification processes for various materials, has been applied to radioactive waste to improve its disposal stability and reduce its volume. Because the thermal decomposition of dry active waste (DAW) significantly reduces its volume, the volume reduction factor of DAW vitrification is high. The KHNP developed the optimal glass composition for the vitrification of DAW. Since vitrification offers a high-volume reduction ratio, it is expected that disposal costs could be greatly reduced by the use of such technology. The DG-2 glass composition was developed to vitrify DAW. During the maintenance of nuclear power plants, metals containing paper, clothes, and wood are generated. ZrO2 and HfO2 are generally considered to be network-formers in borosilicate-based glasses. In this study, a feasibility study of vitrification for DAW that contains metal particulates is conducted to understand the applicability of this process under various conditions. The physicochemical properties are characterized to assess the applicability of candidate glass compositions.

Air conditioner filters purify the air of indoor environments by removing air pollutants and supporting the efficiency of the unit’s cooling function. However, an air conditioner filter can become a microenvironment in which some fungi can grow as dust continues to accumulate and favorable humidity conditions are formed. Fungal growth in air conditioner filters could lead to fungal allergies or fungal diseases, in addition to emitting a foul odor. In an effort to understand what species causes this malodorous problem, we investigated the diversity of fungi found in air conditioners. Fungi were sampled from the collected air conditioner filters and grown on DG18 agar media. After purification for pure isolates, species identification was undertaken. Colony morphology was observed on PDA, MEA, CYA, and OA media. Microstructures of fruiting body, mycelia, and spores were examined using a light microscope. Molecular identification was performed by PCR and sequencing of PCR amplicons, and molecular phylogenetic analysis of sequenced DNA markers, including the Internal Transcribed Spacer (ITS), the 28S large subunit of the nuclear ribosomal RNA (LSU rDNA), the β-tubulin (BenA) gene, the Calmodulin (CaM) gene, and the DNA-directed RNA polymerase II subunit 2 (RPB2) gene. Through this identification process, we found two fungal species, Aspergillus miraensis and Dichotomopilus ramosissimus, which are unrecorded species in Korea. We will now report their morphological and molecular features.

The thermal treatment of radioactive waste attracts great attention. The thermal treatment offers lots of advantages, such as significant volume reduction, hazard reduction, increase of disposal safety, etc. There are various thermal technologies to waste. The developed technologies are calcination, incineration, melting, molten salt oxidation, plasma, pyrolysis, synroc, vitrification, etc. The off-gas treatment system is widely applied in the technologies to increase the safety and operation efficiency. The thermal treatment generates various by-product and pollutants during the process. The dust or fly ash are generated as a particulate from almost every radioactive waste. The treatment of PVC related components generates hydrogen chloride, which usually brings corrosion of facility. The treatment of rubber and spent resin generates sulfur oxide, SOx. The treatment of nitrile rubber generates nitrogen oxide, NOx. The incomplete combustion of radioactive waste usually generates carbon oxide, COx. The process temperature also affects the generation of off gas, such as NOx and/or COx. Various off gas treatment components are organized for the proper treatment of the previously mentioned materials. In this study systematical review on off gas treatment will be reported. Also, worldwide experiences and developed facility will be reported.

The primary purpose of high temperature process of radioactive waste is to satisfy the waste acceptance criteria and volume reduction. The WAC offers the guideline of waste form fabrication process. The WAC is defined as quantitative or qualitative criteria specified by the regulatory body, or specified by and operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. The main objective of WAC is to protect staff and general public and environment by the containment of radioactive material, limit external radiation level, and prevent criticality. The WAC also offers systematic management of radioactive waste by standardization of waste management operations, facilitation waste tracking, ensure safe and effective operation of operating facilities, etc. Since the high temperature process for radioactive waste is considered in many countries, lots of codes and standards are considered. In many WACs, compressive strength, thermal cycle stability, radiation exposure stability, free liquid, and leachability are evaluation to understand the effect of solidified form to the disposal facility. In this paper, systematical review on waste form will be discussed. In addition, brief result of characterization of waste form will be compared.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

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.

During the operation of the nuclear power plant, various radioactive waste are generated. The spent resin, boron concentrates, and DAW are classified as a generic radioactive waste. They are treated and stored at radioactive waste building. In the reactor vessel, different types of radioactive waste are generated. Since the materials used in reactor core region exposed to high concentration of neutrons, they exhibit higher level of surface dose rate and specific activity. And they are usually stored in spent fuel pool with spent fuel. Various non-fuel radioactive wastes are stored in spent fuel pool, which are skeleton, control rod assembly, burnable neutron absorber, neutron source, in core detector, etc. The skeleton is composed of stainless 304 and Inconel-718. There are two types of control rod assembly, that are WH type and OPR type. The WH type control rod is composed of Ag-In-Cd composites. The OPR type control rod is composed of B4C and Inconel-625. In this paper, the characteristics and storage status of the non-fuel radioactive waste will be reported. Also, the management strategy for the various non-fuel radioactive waste will be discussed.

The segmentation of activated components is considered as a one of the most important processes in decommissioning. The activated components, such as reactor vessel and reactor vessel internals, are exposed to neutron from the nuclear fuel and classified to intermediate, low, and very low-level wastes. As it is expected, the components, which are closed to nuclear fuel, exhibit higher degree of specific activity. After the materials were exposed to neutrons, their original elements transform to other nuclides. The primary nuclides in activated stainless steel are 55Fe, 63,59Ni, 60Co, 54Mn, etc. The previous study indicates that the specific activity of individual nuclide is strongly depends on the material compositions and impurities of the original materials. The 59Co is the one of the most important impurities in stainless steel and carbon steel. In this paper, the relationship between individual nuclides in activation analysis of activated components was studied. The systematic study on specific activity of primary nuclides will be discussed in this paper to understand the activation tendency of the components.

Dry active waste (DAW) contains substantial amount of cellulose related materials. The DAW are usually classified as low and/or very low-level waste. In Korea, three types of disposal facilities have been considered: silo, engineering barrier, and land-fill. Currently, only the silo type disposal facility is in operation. Around 27 thousand drums were disposed in silo. Massive amount of cement concrete is used in construction of silo. The ground waste, which flow through the concrete structure, shows higher pH than as it is. It is generally known that the pH of silo is ~12.47 in Korea, when considering construction material, filling material, and property of ground water. It is expected that the cellulose in DAW will be partially transformed to isosaccharinic acid (ISA). It is generally accepted that the ISA plays a negative role in safety analysis of disposal facility by stimulation of specific nuclides. Various factors affect the degradation of cellulose containing radioactive waste, such as degree of polymerization, pH of disposal condition, interaction between concrete structure and ground water, etc. In this paper, the disposal safety analysis of cellulose containing radioactive, usually paper, cotton, wood, etc., are studied. The degradation of cellulose with respect to degree of polymerization, pH of neighboring water, filling material of silo, etc. are reviewed. Based on the review results, it is reasonable to conclude that the substantial amount of DAW could be disposed in silo.

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’ radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

Lysophosphatidic acid (LPA) is a bioactive lipid messenger involved in the pathogenesis of chronic inflammation and various diseases. Recent studies have shown an association between periodontitis and neuroinflammatory diseases such as Alzheimer’s disease, stroke, and multiple sclerosis. However, the mechanistic relationship between periodontitis and neuroinflammatory diseases remains unclear. The current study found that lysophosphatidic acid receptors 1 (LPAR1) and 6 (LPAR6) exhibited increased expression in primary microglia and astrocytes. The primary astrocytes were then treated using medium conditioned to mimic periodontitis through addition of Porphyromonas gingivalis lipopolysaccharides, and an increased nitric oxide (NO) production was observed. Application of conditioned medium from human periodontal ligament stem cells with or without LPAR1 knockdown showed a decrease in the production of NO and expression of inducible nitric oxide synthase and interleukin 1 beta. These findings may contribute to our understanding of the mechanistic link between periodontitis and neuroinflammatory diseases.

During the decommissioning of nuclear power plant (NPP), massive amount of concrete wastes is generated, which are non-radioactive and radioactive. The concrete is a representative construction material which affords reliable structural stability, good formability, and trustful integrity. Also, its reasonable neutron absorbing property allows the various application for many components, including building construction material, bio-shield concrete, etc. Due to the noted aspects of concrete, the radiological concrete characterization is classified as an important process for development of effective strategy for concrete management, in terms of process management and financial control during the decommissioning. The characterization of bio-shield concrete is important in waste management. The understanding and characterization of activation depth is essential for the determination of waste management strategy, classification of bio-shield concrete, and process development of decommissioning. On the other hand, concrete for construction application requires the evaluation of surface contamination of them. The concrete for containment building and its structure is rarely activated, but surface contaminated. In this paper, the reactor data from representative PWR reactors in the US is studied. The experience on Yankee Rowe, Maine Yankee, and Connecticut Yankee NPPs are systematically studied. The Yankee Rowe was a 4-loop PWR of Westinghouse design with 185 MWe. The Main Yankee was a 3- loop PWR of Combustion Engineering design with 864 MWe. The Connecticut Yankee was a 4-loop Westinghouse type with 560 MWe. The characterization studies on bio-shield concrete will be discussed in this paper, including activation depth, primary nuclides, etc.