This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

Currently, non-volatile nuclides such as 94Nb, 99Tc, 90Sr, 55Fe, and 59/63Ni are used a sequential separation. In this study, we developed a separation for 99Tc and 90Sr by a carbonate precipitation. Sodium Carbonate (Na2CO3) was inserted in the aqueous sample from a Dry Active Waste (DAW) and a carbonate precipitation was produced. The precipitate is composed of di- or tri-valent element such as Co, Sr, Fe, Ni and the supernatant is composed of mono-valent element (Cs) and anion materials (ReO4 -, TcO4 -). In DAW, it was confirmed that the recovery of 90Sr (precipitate) and 99Tc (supernatant) were > 90%, respectively. The precipitate and supernatant separated by using a Sr-resin and an anion-exchange resin, respectively. The separated samples were measured by a Liquide Scintillation Counter (LSC, 90Sr) and Induced-Coupled Plasma-Mass Spectroscopy (ICPMS, 99Tc).

Bis (2-ethylhexyl)phosphoric acid (HDEHP) is a renowned extractant, favored for its affinity to selectively remove uranium via its P=O groups. We previously synthesized HDEHP-functionalized mesoporous silica microspheres for solid-phase uranium adsorption. Herein, we investigated the kinetic and isothermal behavior of uranyl ion adsorption in mesoporous silica microspheres functionalized with phosphate groups. Adsorption experiments were conducted by equilibrating 20 mg of silica samples with 50 mL of uranium solutions, with concentrations ranging from 10 to 100 mgU L−1 for isotherms and 100 mgU L−1 for kinetics. Three distinct samples were prepared with varying HDEHP to TEOS molar ratios (x = 0.16 and 0.24) and underwent hydrothermal treatment at different temperatures, resulting in distinct textural properties. Contact times spanned from 1 to 120 hours. For x = 0.16 samples, it took around 50 and 11 hours to reach equilibrium for the hydrothermally treated samples at 343 K and 373 K, respectively. Adsorbed quantities were similar (99 and 101 mg g-1, respectively), indicating consistent functional group content. This suggests that the key factor influencing uranium adsorption kinetics is pore size of the silica. The sample treated at 373 K, with a larger pore size (22.7 nm) compared to 343 K (11.5 nm), experienced less steric hindrance, allowing uranium species to diffuse more easily through the mesopores. The data confirmed the excellent fit of pseudo-second-order kinetic model (R2 > 0.999) and closely matched the experimental value, suggesting that chemisorption governs the rate-controlling step. To gain further insights into uranium adsorption behavior, we conducted an adsorption isotherm analysis at various initial concentrations under a constant pH of 4. Both the Langmuir and Freundlich isotherm models were applied, with the Langmuir model providing a superior fit. The relatively high R2 value indicated its effectiveness in describing the adsorption process, suggesting homogenous sorbate adsorption on an energetically uniform adsorbent surface via a monolayer adsorption and constant adsorption site density, without any interaction between adsorbates on adjacent sites. Remarkably, differences in surface area did not significantly impact uranium removal efficiency. This observation strongly suggests that the adsorption capacity is primarily governed by the loading amount of HDEHP and the inner-sphere complexation with the phosphoryl group (O=P). Our silica composite exhibited an impressive adsorption capacity of 133 mg g-1, surpassing the results reported in the majority of other silica literature.

The first commercial operation of Kori-1, which commenced in April 1978, was permanently shut down in June 2017, with plans for immediate dismantling. The decommissioning process of nuclear power plants generates a substantial amount of radioactive waste and poses significant radiation exposure risks to workers. Radioactivity is widely distributed throughout the primary coolant system of the reactor, including the reactor pressure vessel (RPV), steam generator (SG), and pressurizer. In particular, the SG has a considerable size and complex geometry, weighing approximately 326 tons and having a volume of 400 m3. The SG tubes are known to contain high levels of radioactivity, leading to significant radiation exposure to workers during the dismantling process. Therefore, this study aims to evaluate the workers’ radiation exposure during the cutting of SG tubes, which account for approximately 95% of the total radiation dose in the SG. Firstly, the CRUDTRAN code, developed to predict the behavior of soluble and particulate corrosion products in a PWR primary coolant system, is used to estimate the radioactive inventory in the SG tubes. Based on decontamination factors (DF) obtained in the SG tubes through overseas experience, the expected reduction in radioactivity during the Kori-1 reactor’s full-system decontamination (FSD) process is considered in the CRUDTRAN results. These results are then processed to estimate the radioactivity in both the straight and bent sections of the tubes. Subsequently, these radioactivity values are used as inputs for the MicroShield code to calculate the worker radiation exposure during the cutting of both straight and bent sections of the tubes. The cutting process assumes that each SG tube section is cut in a separate, shielded area, and the radiation exposure is assessed, taking into account factors such as cutting equipment, cutting length, working hours, and working distance. This study evaluates the worker radiation exposure during the cutting of SG tubes, which are expected to have a significantly high radioactivity due to chalk river unidentified deposit (CRUD). This assessment also considers the reduction in radioactivity within the steam generator tubes resulting from the FSD process. Consequently, it enables an examination of how worker radiation exposure varies based on the extent of FSD. This study may provide valuable insights for determining the scope and extent of the FSD process and the development of shielding methods during the dismantling of SG tubes in the future.

The nuclear licensee must ensure that the nuclear or radiological emergency preparedness and response organization is explicitly defined and staffed with adequate numbers of competent and assessed personnel for their roles. This paper describes the responsibilities of medical and support personnel for the medical action of casualties in the event of a radiological emergency at the KAERI. Currently, there is one medical personnel (nurse) in KAERI, and a total of eight medical support personnel are designated for medical response in the event of a radiological emergency. These medical support personnel are designated as one or two of the on-site response personnel at each nuclear facility, operating as a dedicated team of A, B (4 people each). In the event of a radiological emergency, not all medical support personnel are mobilized, but members of the dedicated medical team, which includes the medical support personnel of the nuclear facility where the accident has occurred, are summoned. Medical and support personnel will first gather in the onsite operational support center (OSC)/technical support center (TSC) to prepare and stand by for the medical response to injured when a radiological emergency is declared. They should take radiation protective measures, such as wearing radiation protective clothing and dosimeters, before entering the onsite of a radiological emergency, because injuries sustained during a radiological emergency may be associated with radioactive contamination. In the event of an injury, direct medical treatment such as checking the patient’s vitals, first aid, and decontamination will be carried out by medical personnel, while support personnel are mainly responsible for contacting the transfer hospital, reporting the patient’s condition, accompanying the ambulance, filling out the emergency medical treatment record, and supporting medical personnel. In order to respond appropriately to the occurrence of injuries, we regularly conduct emergency medical supplies education and medical training for medical support personnel to strengthen their capabilities.

As nuclear decommissioning ventures become increasingly complex, the role of digitalization in facilitating and enhancing these operations is becoming indispensable. This transition to a more digitized approach presents a myriad of advantages, including: augmented avenues for data acquisition, analysis, and visualization to bolster dismantling strategies; simulations in virtual environments for operator training; precise forecasting of future waste emergence, culminating in refined cost estimations; and more immersive decommissioning visualizations for both operators and external stakeholders. Salient benefits conferred by the integration of digital technologies in decommissioning encompass improved collaboration, enriched knowledge transfer, clarity regarding present technological constraints, insights into key influencing factors, clearer criteria for technology selection, and a profound understanding of the potential challenges and merits of a broader incorporation of digital tools in decommissioning endeavors. Of paramount importance is the opportunity presented for superior workforce training and safety measures, exemplified by ALARAbased planning. Amidst the myriad facets of digital adoption, 3D modeling of nuclear facilities derived from laser-scanned point clouds stands out as a pivotal domain in the digitalization. The transformation of intricate point cloud data into a comprehensible 3D mesh remains the crux of this paper. The process of mesh generation, despite being simpler than its counterpart of converting to a 3D solid model, is crucial for multiple reasons. The resultant 3D mesh offers an enhanced visual representation compared to a sparse point cloud, paving the way for improved spatial perception. Furthermore, it serves as a rudimentary tool for approximating component volumes and the ensuing waste, thereby playing an instrumental role in waste manipulation strategies, notably in collision detection. This paper delves deep into the nuances of mesh generation, conducting an parametric study of mesh conversion algorithms, including down-sampling rates. Through this rigorous examination, we endeavor to shed light on optimal methodologies, hoping to catalyze advancements in the digitalization of nuclear decommissioning processes.

KEPCO KPS is the contractor for the full system decontamination (FSD) of Kori Unit 1 and under preparation such as modification, lay out for equipment installation, setting up tie-in/out point for chemical injection and way to pressurize the system, of its successful performance. In this research, KPS introduced how KPS has designed and prepared for the FSD project and how will the chemical decontamination process be implemented. As described in the previous research, chemical decontamination process is planned to be conducted for three cycles and each cycle is consisted of oxidation, reduction, decomposition, and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of IX by the heat. If the decontamination result does not meet the target DF and the dose rate, additional cycle can be conducted. Expected volume of process water for FSD is 200 m3. Three systems have been designated as decontamination targets: reactor coolant system (RCS), residual heat removal system (RHRS), chemical volume control system (CVCS). For the steady flow rate, existed plant equipment such as reactor coolant pump (RCP) will be operated and modifications on some components will be conducted. Due to the limited space for installation, decontamination equipment and other resources are distributed to three different places. KPS designed the layout of equipment installed inside the containment vessel. The layout contains the information of shielding for highly radiated equipment such as IX and filter skid.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for fuel examinations. The facility has pools and hot cells for handling and examining fuel assemblies and rods. Among the hot cells, the second cell is for measuring rod internal pressure (RIP) and then cutting the rod to make samples for destructive tests. Currently, the cutting machine is broken, so it has to be replaced. Because the existing cutting machine consists of many parts and its size was quite a bit large to handle and treat for the radioactive waste disposal, the disassembly work has been performed to make it smaller using manipulators. The drawings of the cutting machine were reviewed and the disassembly tools were developed considering workability when the work performed at the hot cell using the manipulators. The large parts such as motor, mirror and cable, etc., were able to be disassembled and the machine size became so smaller that it could be easily handled for the disposal.

As the acceptance criteria for low-intermediate-level radioactive waste cave disposal facilities of Korea Radioactive Waste Agency (KORAD) were revised, the requirements for characterization of whether radioactive waste contains hazardous substances have been strengthened. In addition, As the recent the Nuclear Safety and Security Commission Notice (Regulations on Delivery of Low- Medium-Level Radioactive Waste) scheduled to be revised, the management targets and standards for hazardous substances are scheduled to be specified and detailed. Accordingly, the Korea Atomic Energy Research Institute (KAERI) needs to prepare management methods and procedures for hazardous substances. In particular, in order to characterize the chemical requirements (explosiveness, ignitability, flammability, corrosiveness, and toxicity) contained in radioactive waste, it must be proven through documents or data that each item does not contain hazardous substances, and quality assurance for the overall process must be provided. In order to identify the characteristics of radioactive waste that will continue to be generated in the future, KAERI needs to introduce a management system for hazardous substances in radioactive waste and establish a quality assurance system. Currently, KAERI is thoroughly managing chelates (EDTA, NTA, etc.), but the detailed management procedures for hazardous substances related to chemical requirements in radioactive waste in the radiation management area specified above are insufficient. The KAERI’s Laboratory Safety Information Network has a total periodic regulatory review system in place for the purchase, movement, and disposal of chemical substances for each facility. However, there is no documents or data to prove that the hazardous substances held in the facility are not included in the radioactive waste, and there are no procedures for managing hazardous substances. Therefore, it is necessary to establish procedures for the management of hazardous substances, and we plan to prepare management procedures for hazardous substances so that chemical substances can be managed according to the procedures at each facility during preliminary inspection before receiving radioactive waste. The procedure provides definitions of terms and types of management targets for each characteristic of the chemical requirements specified above (explosiveness, ignition, flammability, corrosiveness, and toxicity). In addition, procedure also contains treatment methods of radioactive waste generated by using hazardous substances and management methods of in/out, quantity, history of that substances, etc. As the law is revised in the future, management will be carried out according to the relevant procedures. In this study, we aim to present the hazardous substance management procedures being established to determine whether radioactive waste contains hazardous substances in accordance with the revised the notice and strengthened acceptance criteria. Through this, we hope to contribute to improving reliability so that radioactive waste could be disposed of thoroughly and safely.

Republic of Korea is preparing to decommission Kori Unit 1 and Wolsong Unit 1. Decommissioning of a nuclear power plant proceeds in the following stages: shutdown, transition period, decontamination, cutting, waste treatment, and site restoration. When nuclear power plant is decommissioned, It is expected that approximately 80,000 drums of radioactive waste will be generated per nuclear power plant. Therefore, various technologies are being researched and developed to reduce this to approximately 14,500 drums. Technologies for waste volume reduction are largely mechanical and electrical/thermal methods. Representative examples of mechanical volume reduction technologies include super compactors and electrical/thermal volume reduction technologies include induction and plasma torch furnaces. Both technologies are effective reduction technologies, but the reduction ratio varies depending on the type or condition of waste before treatment. For example, as a result of testing waste reduction using a super compactor at NUKEM in Germany, the reduction ratio was found to be between 1.3 and 7 depending on the type or condition of waste such as chips, ash, scrap metal, sand, etc. And according to IAEA-TECDOC-1527, when reducing the volume of metals, aluminum, lead, copper, brass, etc. using induction melting, the waste volume reduction ratio is 5 to 20. In this paper, referring to these results, a melting test was conducted using a previously developed plasma torch with an output of more than 100 kW. And volume reduction characteristics of this plasma torch was considered depending on waste type or condition.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.