Once decommissioning begins, it is expected that large amount of radioactive wastes will be produced in a short period of time. The expected amount of radioactive wastes from Kori unit 1 NPP are approximately 80,000 drums (base on 200 L). By minimizing the amount of radioactive wastes generated through decontamination and reduction, KHNP has set the final target for the amount of radioactive wastes to be delivered to the disposal site at approximately 14,500 drums. Here, plasma torch melting technology is an essential technology for radioactive wastes treatment during nuclear power plants decommissioning and operation, because of its large volume reduction effects and the diversity of disposable wastes. KEPCO KPS was able to secure experience in operating Plasma Torch Melter (PTM) by conducting a research service for ‘development of plasma torch melting system advancement technology’ at KHNP-CRI. This study will compare kilo and Mega-Watt class PTM, largely categorized into facility configurations, operating parameters, and waste treatment. Based on this study, it would be desirable to operate PTM with approximate capacity according to the frequency and amount of waste production, and suggest volume for a kilo and Mega-watt class plasma torch in the melting furnace respectively. This plays to its strengths for both a kilo and Mega-watt class PTM.

To safely dispose of highly radioactive spent resin and concentrate waste generated through nuclear power plant operations, it is essential to meet the physicochemical properties requirements of the packages and ensure the accuracy and reliability of radiological characteristics determination. Both spent resin and concentrate are packaged in high-integrity containers (HICs) after drying and are homogeneous waste products generated in the primary system and liquid radioactive waste treatment system. Meeting the physicochemical properties requirements does not appear to be difficult. However, to achieve reliable radiological characterization of high-integrity container packages, it is necessary to take a representative sample and perform accurate radiological analysis. Therefore, this paper discusses the methodology for evaluating the radionuclide inventory of high radioactive resin and concentrate packages, as well as the essential element technology and considerations. For relatively high radioactive resin and concentrate packages, the radionuclide inventory for each package should be evaluated with high reliability through direct radiological analysis of the representative samples collected for each package. This can contribute to the efficient operation of radioactive waste disposal facilities. Radionuclide-specific concentrations directly analyzed for each package will be managed in a database. As analytical data accumulates and direct measurements of high-integrity container package such as the radwaste drum assay system (RAS) become feasible, statistical techniques such as correlation analysis between easy-tomeasure (ETM) nuclides and difficult-to-measure (DTM) nuclides can lead to the development of efficient and reasonable indirect evaluation methods, such as scaling factor and the mean activity concentration method. As for the element technology, a remote representative sampling technique should be developed to safely and effectively take representative samples of highly radioactive materials, including granulated or hardened concentrate waste. Considerations should also be given to determining the sample quantity representing each package, as well as establishing radiation calibration and measurement methods appropriate to the radiation levels of the representative samples.

In general, radioactive waste with high radioactivity is made into a solid form with performance such as leaching restriction, shape retention, and structural stability so that radioactive waste does not affect humans and the environment as much as possible. This should be applied equally to radioactive waste, whether homogeneous or heterogeneous. The requirements are stipulated in the “Low and Intermediate Level Radioactive Waste Delivery Regulations” notice of the Korea Nuclear Safety and Security Commission. On the other hand, the waste acceptance criteria for domestic disposal facilities require immobilization of heterogeneous waste when the activity concentration is above a certain level, but do not provide specific immobilization performance requirements. In this study, the immobilization requirements applied to heterogeneous radioactive waste in various overseas countries operating low and intermediate-level radioactive waste disposal facilities were studied. First, the IAEA’s safety standards for radioactive waste immobilization, domestic regulations, and disposal facility waste acceptance criteria were reviewed. Countries operating surface disposal facilities such as the United States, France, Spain, and Japan and countries operating underground disposal facilities such as Sweden and Finland were divided to review the current status of immobilization application to heterogeneous waste in overseas countries. When reviewing overseas cases, each country’s disposal methods, types of disposal waste, and waste treatment criteria were also reviewed. It was found that the immobilization requirements for heterogeneous radioactive waste vary depending on the disposal method and the type of barrier used to ensure disposal safety in each country. The common point is to surround heterogeneous radioactive waste within a concrete lining of a certain thickness, and to apply the thickness, compressive strength, and diffusion coefficient of the concrete lining as immobilization performance requirements. Through this study, the immobilization performance requirements for heterogeneous radioactive waste in various overseas countries that stably operate low- and intermediate-level radioactive waste were confirmed, which is expected to contribute to specifying the performance requirements for immobilization of heterogeneous radioactive waste in domestic disposal facilities.

Among domestic Nuclear Power Plants (NPPs), there are a total of 10 nuclear power plants whose operating license expires by 2030, excluding Kori unit 1 and Wolsong unit 1, which are permanently shut downed. Continued operation of these nuclear power plants is being reviewed as a government task. For continued operation, nuclear power plant owners must prepare periodic safety review and other evaluation reports to receive reviews to maintain safety even during continued operation. In the safety evaluation of NPP, it is important to refer to overseas cases and operation experiences. In this study, the matters of radioactive waste management for continued operation of NPP was considered by analyzing the safety evaluation reports and safety enhancements of license renewal of NPP in USA, Radioactive waste generated from NPPs can be classified into solid, liquid, and gaseous states. Radioactive waste generated during the operation and maintenance of power plants is classified, stored and treated in the radioactive waste management system according to the source. Equipment and monitors related to radioactive waste management are continuously operated, managed, inspected according to standards and maintain their original functions. Various activities to reduce the generation and emission of radioactive waste from NPPs are performed. After reviewing the NRC’s safety evaluation report on the application documents for license renewal of US NPPs (Sequoyah, Byron and braidwood) the evaluation details and matters requiring enhancement for the radioactive waste management system were confirmed. As a major check, selective leaching occurred in the body of the gray cast iron valve and the heat exchanger shell containing the copper alloy exposed to the radioactive waste liquid. Selective leaching causes loss of material and may interfere with the original function of the facility, so management is required. For the safe operation and management of NPPs, it is important to refer to overseas cases and experiences. Among the safety evaluations for the continued operation of domestic NPPs, in the field of the radioactive waste management system, if the case of the US NPP is referred to, the review by the regulatory body and the action taken by the licensee will be more efficient.

Level measurement of liquid radwaste is essential for inventory management of treatment system. Among various methods, level measurement based on differential pressure has many advantages. First, it is possible to measure the liquid level of the system regardless of liquid type. Second, as the instrument doesn’t need to be installed near the tank, there is no need to contact the tank when managing it. Therefore, workers’ radiation dose from the system can be decreased. Finally, although it depends on the accuracy, the price of the instrument is relatively low. With these advantages, in general, liquid radwaste level in a tank is measured using differential pressure in the treatment system. Not only the advantages described above, there are some disadvantages. As the liquid in the system is waste, it is not pure but has some suspended materials. These materials can be accumulated in tanks and pipes where the liquids move to come into direct contact with pneumatic pipes that are essential in differential pressure instruments. As a result, in case of a treatment using heat source, the accumulated materials may become sludge causing interference in pneumatic pipes. And this can change the pressure which also affects the level measured. In conclusion, in case of liquid storage tanks in which the situation cannot be checked, the proficiency of an operator becomes important.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.

The purpose of full system decontamination before decommissioning a nuclear power plant is to reduce radiation exposure of decommissioning workers and to reduce decommissioning waste. In general, full system decontamination removes the CRUD nuclides deposited on the inner surface of the reactor coolant system, chemical and volume control system, residual heat removal system, pressurizer, steam generator tube, etc. by chemical decontamination method. The full system decontamination process applied to Maine Yankee and Connecticut Yankee in the USA, Stade, Obrigheim, Unterweser, Nekawestheim Unit 1 in Germany, Mihama Unit 1 and 2 in Japan, Jose Cabrera Unit 1 in Spain, and Barseback Unit 1 and 2 in Sweden are HP/CORD UV, NP/CORD UV, and DfD. In this study, the quantity of 60Co radioactivity removal, metal removal, ion exchange resin and filter generation according to reactor power, surface area and volume of the full system decontamination flow path, and the decontamination process were compared and analyzed. In addition, the quantity of 60Co radioactivity removal by each nuclear power plant was compared and analyzed with the evaluation results of the 60CO radioactivity inventory of the Kori Unit 1 full system decontamination loops conducted by SAE-AN Enertech Corporation.

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

After the Fukushima accident in 2011, a huge amount of radioactively contaminated water is being generated by cooling the melted fuel of units 1, 2 and 3. Most of contaminated water is seawater and underwater containing not only salt elements but also nuclear fission products with radioactivity. To treat the contaminated water, Cs/Sr removal facilities such as KURION and SARRY are being operated by TEPCO. Additionally, three ALPS facilities are on operation to meet the regularity standards for discharge to the sea. However, massive secondary wastes such as Zeolite, sludge and adsorbent is being generated by these facilities for liquid water treatment. The secondary wastes containing various radionuclide with Cs and Sr is difficult to store due to highly radioactive concentration and corrosive properties. In Japan, a variety of technologies such as GeoMelt vitrification, In-Can vitrification and CCIM vitrification is considered as a promising solution. In this study, they were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for thermal treatment of sludge radwaste.

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.

During the operation or decommission of nuclear facilities, a large amount of dry active waste and cable waste with various shape and material is generated. Most of these wastes have almost no radioactive contamination and can be disposed of by incineration, landfill, recycling, etc. under clearance regulation. For clearance of radioactive waste, it is necessary to verify the characteristics of radiological contamination and prove that it meets the criteria for clearance regulation. According to the domestic clearance regulation, if it is difficult to measure radioactivity of wastes due to their surface condition using direct or indirect measurement methods, representative samples should be collected and analyzed for radioactivity. When sampling, it is desirable to collect samples of about 1 kg that can represent waste contamination per 200 kg or per 1 m2, and the homogeneity of the samples also should be demonstrated. However, in the case of dry active wastes, it is very difficult to prove the homogeneity of the samples because of surface shapes and conditions of the wastes. In particular, considering cable waste generated during the decommission, it is hardly capable to prove the representativeness of the sample, even though the inner shell of the covering material and the copper wire are almost uncontaminated. In this study, we show the development of a treatment system that makes it easy to prove the representativeness of samples when disposing of dry active waste or cable waste generated in nuclear facilities. The treatment device is designed in such a way that it has different storage unit and cutting unit suitable for the material characteristics of each waste type (soft, hard and cable), and therefore optimizes the efficiency of the shredding or cutting process. In addition, it is expected that the work efficiency in the radioactive treatment site with a narrow area can also be improved by providing a moving part on the device.