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.

Licensing for the application of the Polymer Concrete High Integrity Container (PC-HIC) to nuclear power plants has been completed or is in progress. Approval for the expanded application to all domestic nuclear power plants has been completed to utilize the 860 L PC-HICs for the 2nd stage surface repository, and the regulatory body is reviewing the license application to use the 510 L PCHICs for the 1st stage underground repository in the representative nuclear power plants. The 860 L PC-HICs, which have been licensed for all domestic nuclear power plants, will be used for safe storage management and disposal of low-dose dried concentrate waste and spent resin, and a total of 100 units is expected to be supplied to representative nuclear power plants that have been licensed first. The 510 L PC-HICs are planned to be used for underground disposal of high-dose spent resin and dried concentrate waste. Prior to the application of PC-HICs to nuclear power plants and disposal to the repository, it is necessary to establish realistic and reasonable requirements through close consultations between waste generator and disposal operators to ensure the suitability for disposal of PC-HIC packages and to carry out disposal delivery and acceptance work. Since the Polymer Concrete High Integrity Container (PC-HIC) has long-term integrity of more than 300 years and the barrier does not temporarily collapse, spent resin and dried concentrate waste, which are radioactive wastes to be solidified, can be disposed of much more safely in PC-HIC packages than solidified types. Acceptance criteria for the PC-HIC packages should be prepared fully reflecting the advantages of PC-HIC, and quality assurance methods for physical/chemical/radiological characterization results based on the Waste Certification Program (WCP) should be supported. In addition, infrastructure should be secured for safe transportation, handling, and storage of the PC-HIC packages. In this paper, we have tried to find a reasonable acceptance criteria, quality assurance method, and infrastructure level according to the dose and disposal conditions of PC-HIC packages.

Domestic nuclear power plants developed the scaling factors for the radioactive waste generated from 2004 to 2008 for the first time. Afterwards, the effectiveness of continuous application of the scaling factors have been evaluated for newly generated radioactive waste over the past two years. It was confirmed that most of the initially developed scale factors were effective within a factor of 10 principle, which is an acceptable range. The scaling factors were updated using the analysis data base from 2004 to 2016. A scaling factor refers to the calculated abundance ratio between Key (Easy-to- Measure) and DTM (Difficult-to-Measure) nuclide at the time of generation of radioactive waste based on the source term in the reactor of an operating nuclear power plant. The effectiveness of continuous application of scaling factors can be evaluated at regular intervals regardless of operation status or when there are events that change scaling factors during nuclear power plant operation, such as zinc injection, large-scale facility replacement, and long-term shutdown etc. Even in the case of a permanently shut down nuclear power plant in which fuel is withdrawn from the reactor and generation of new nuclides by nuclear fission and radiation has stopped, periodic verification is conducted to confirm whether the scaling factor developed before permanent shutdown can be effectively applied to the radioactive waste generated after permanent shutdown. However, depending on the nuclear power plant decommissioning strategy or conditions, the period of permanent shutdown prior to decommissioning can be very long, so preparations are needed to ensure the appropriateness of scaling factor operation. In the case of domestic nuclear power plants, Kori Unit 1, a light water reactor, was permanently shut down in June 2017, and as a heavy water reactor nuclear power plant, the permanent shutdown of Wolseong Unit 1 was finally decided in December 2019 after twists and turns including large-scale facility replacement and long-term shutdown. In this paper, we have predicted when the scaling factors will change significantly due to radioactive decay and the difference in halflife between the Key and DTM nuclides over time after permanent shutdown. We also have tried to find appropriate countermeasures for the operation of scaling factors during permanent shutdown period, such as updating scaling factors or applying correction factors.

Radioactive spent resin and concentrate waste powder generated from the primary system of nuclear power plants (NPPs) should be treated and disposed of in the form of solidified products or high integrity container (HIC) packages. We are preparing for the application of polymer concrete high integrity containers (PC-HICs) that has been approved for disposal and field application after going through the disposal suitability review of the repository operator and the license review process of the regulatory body. A reliable assessment of nuclide inventory in waste drum is required for the disposal of the radioactive waste drums, and the representative samples should be collected for both the indirect (non-destructive assessment based on the scaling factor, average radioactivity concentration, etc.) and direct (destructive analysis) evaluation of the difficult-to-measure (DTM) nuclides. It is important to secure the representativeness of samples for reliable and accurate evaluation of radionuclide inventory and approval of methodologies for highly radioactive waste such as spent resin and concentrate waste poser, and in order to secure the radiation safety of the sampling workers and representativeness of the samples, a remote sampling method is required with excellent convenience and safety and sufficient representativeness of the sample. The simple sampling device used in the past to collect samples for the scaling factor does not have a remote control function, so high-radiation samples must be collected within a very short time and it is difficult to obtain sufficiently representative samples due to structural characteristics that cannot collect the entire sample in the axial direction of the package. Therefore we developed concept designs for a remote sampling device that can satisfy both sample representativeness, operator convenience and safety.

High Integrity Container (HIC) made of polymer concrete was developed for the efficient treatment and safe disposal of radioactive spent resin and concentrate waste in consideration of the disposal requirements of domestic disposal sites. Permission for application of Polymer Concrete High Integrity Container (PC-HIC) to the domestic nuclear power plants (NPPs) has been completed or is under examination by the regulatory agency. In the case of 860 L PC-HIC for very-low-level-waste (VLLW) or low low-level-waste (LLW), the application of four representative NPPs has been approved, and the license for extended application to the rest NPPs is also almost completed. A licensing review is also underway to apply 510 L PC-HIC for intermediate and low-level-waste (ILLW) to representative nuclear power plants. In order to handle and efficiently store and manage PC-HICs and high-dose PCHIC packages, a gripper device that can be remotely operated and has excellent safety is essential, and the introduction of NPPs is urgent. The conventional gripper device developed by the PC-HIC manufacturer for lifting test to evaluate the structural integrity of PC-HIC requires a rather wide storage interval due to its design features, and does not have a passive safety design to handle heavy materials safely. In addition, work convenience needs to be reinforced for safety management of high radiation work. Therefore, we developed a conceptual design for a gripper device with a new concept to minimize the work space by reflecting on-site opinions on the handling and storage management conditions of radioactive waste in NPPs, and to enhances work safety with the passive safety design by the weight of the package and the function of checking the normal seating of the device and the normal operation of the grip by the detector/indicator, and to greatly improves the work efficiency and convenience with the wireless power supply function by rechargeable battery and the remote control function by camera and wireless monitoring & control system. Through design review by experts in mechanical system, power supply and instrumentation & control fields and further investigations on the usage conditions of PC-HICs, it is planned to facilitate preparations for the application of PC-HIC to domestic NPPs by securing the technical basis for a gripper device that can be used safely and efficiently and seeking ways to introduce it in a timely manner.

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.

In order to indirectly evaluate the inventory of difficult-to-measure (DTM) nuclides in radioactive waste, the scaling factor method by key nuclide has been used. It has been usually applied to low-and intermediate-level dry active waste (DAW), and the tolerance of 1,000% margin of error in the US, that is the factor of 10, is applied as an allowable confidence limits considering the inhomogeneity of the waste and the limitation of sample size. This is because the scaling factor method is based on economic efficiency. Confidence limits is the uncertainty (sampling error) according to predicting the mean value of the population by the mean value of the sample at 95% confidence level, reflecting the limitations of sample size (representation) with the standard deviation. If the standard deviation is large, the sample size can be increased to satisfy the allowable confidence limits. In the new nuclear power plants, the concentration of cesium nuclide (137Cs) in radioactive waste tends to be very low due to advances in nuclear fuel and reactor core management technology, which makes it very difficult to apply cesium as a key nuclide. In addition, it is inevitable to apply the mean activity concentration method, which reasonably and empirically derives the concentration of DTM nuclides regardless of key nuclide, when the correlation between key and DTM nuclides is not significant. The mean activity method is a methodology that applies the average concentration of a sample set to the entire population, and is similar to applying the average concentration ratio between key and DTM nuclides of a sample set to the population in the scaling factor method. Therefore, in this paper, the maximum acceptable uncertainty (confidence limits) at a reasonable level was studied when applying the mean activity concentration method by arithmetic mean unlike the scaling factor method which usually uses the geometric mean method. Several measures were proposed by applying mutatis mutandis the acceptable standard deviation in radiation measurement and the factor of 10 principle, etc., and the appropriateness was reviewed through case analysis.

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.