In 2017, a decision was made to permanently shut down Kori Unit 1, and preparations began to be made for its decontamination and decommissioning. The dismantling of the biological shields concrete, reactor vessel (RV), and reactor vessel internals (RVI) is crucial to the nuclear decommissioning process. These components were radiologically activated by the neutron activation reaction occurring in the reactor during its operational period. Because of the radioactivity of the RV and RVI of Kori Unit 1, remotely controlled systems were developed for cutting within the cavity to reduce radiation exposure. Specialized equipment was developed for underwater cutting operations. This paper focuses on modeling related to RVI operations using the MAVRIC code and the dose calculation for a diver entering the cavity. The upper and lower parts of the RVI are classified as low-level radioactive waste, while the sides that came into contact with the fuel are classified as intermediate-level radioactive waste. Therefore, the modeling presented in this paper only considers the RVI sides because the upper and lower parts have a minimal impact on the radiation exposure. These research findings are anticipated to contribute to enhancing the efficiency and safety of nuclear reactor decommissioning operations.

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

Molten Salt Reactor (MSR) is one of the 4th generation nuclear power systems which is its verified technology in physically and chemically. Among the various salts used for MSR system, the eutectic composition of NaCl-MgCl2 system maintains the liquid state at around 450°C, in the same time, it has high solubility for nuclear fuel chlorides. This characteristic has high advantage for lowering the operating temperature for the MSR, which could reduce the problem of hightemperature corrosion by salt for structural materials significantly. In particular, since MgCl2 has the similar standard reduction potential with nuclear fuel, is used as a surrogate for, many basic researches have been conducted for verifying characteristic of MgCl2. It is well-known that main short-advantage of MgCl2 is hygroscopic properties. MgCl2 changes to MgCl2-xH2O state easily by absorbing moisture in air condition. The hydrated MgCl2 is producing MgOHCl by thermally decomposing at high temperature, the formed MgOHCl corrodes structural materials, even small amount of MgOHCl gives significant damage. Therefore, the purification of MgCl2 has been required for long-term operation of MSR using MgCl2 as a base salt. In this study, the purification of eutectic composition salt for NaCl-MgCl2 has been mainly performed by considering its thermodynamic properties and electrochemical characteristic, and the experimental results have been discussed.

Heavy water primary system decontamination technology is essential to reduce worker exposure and improve safety during maintenance and decommissioning of nuclear facilities. Advanced decontamination technology development aims to secure controlled decontamination technologies that can reduce the cost of radiation exposure and dramatically reduce the amount of secondary waste generated when decontaminating large equipment and large-area facilities. We conducted a study to identify candidate corrosion inhibitors through the literature and analyze the degree of corrosion of carbon steel samples. Countries with advanced nuclear technology have developed chemical decontamination technology for the entire nuclear power generation system and applied it to the dismantling and maintenance of nuclear power plants. In the decontamination process, the corrosion oxide film must be removed. If the base metal is corroded by the decontaminant in this process, additional secondary waste is generated and treatment costs increase. Therefore, it is necessary to develop a corrosion inhibitor that inhibits the corrosion of the carbon steel base metal in the decontamination process to generate a secondary waste liquid that is favorable for waste reduction and treatment. In this presentation, a study was conducted to analyze the extent of corrosion on a carbon steel base material and identify candidate materials for corrosion inhibition testing. Samples were analyzed using optical microscopy and EPMA analysis to determine the thickness of the corroded oxide film. EPMA analysis also allowed us to map the elemental distribution of the carbon steel corrosion layer, which we plan to quantify in the future. The candidate materials for organic-based corrosion inhibitor were also selected based on their inhibition mechanism; having high electronegative elements for coordinate covalent bonding at metal surface and hydrophobic nonpolar group for preventing access of corrosive substances.The selection of candidate materials for corrosion inhibition testing was based on the mechanism of the corrosion inhibitor. Organic-based corrosion inhibitors are adsorbed by donor-acceptor interactions between metal surfaces and highly electronegative elements. Corrosion can also be inhibited by arranging hydrophobic nonpolar groups on metal surfaces in the solution direction to prevent access of corrosive substances.

The primary heat transport system consists mainly of the in-core fuel channels connected to the steam generators by a system of feeder pipes and headers. The feeders and headers are made of carbon steel. Feeders run vertically upwards from the fuel channels across the face of the reactor and horizontally over the refueling machine to the headers. Structural materials of the primary systems of nuclear power plants (NPPs) are exposed to high temperature and pressure conditions, so that the materials employed in these plants have to take into accounts a useful design life of at least 30 years. The corrosion products, mainly iron oxides, are generated from the carbon steel corrosion which is the main constituent of the feeder pipes and headers of this circuit. Typical film thickness on CANDU-PHWR surface is 75μm or 30mg/cm2. Deposits on PHWR tends to be much thicker than PWR due to use of carbon steel and also for the source of corrosion products available on the carbon steel surface. Degradation of carbon steel for the feeder pipes transferring the primary system coolant by flow-assisted corrosion in high temperature has been reported in CANDU reactors including Point Lapreau, Gentully-2, Darlington and Bruce NPPs. The formation of Fe3O4 film on a carbon steel surface reduces the dissolution rate of steel substantially. The protectiveness of the Fe3O4 film over the carbon steel is affected by the environmental factors and the operational parameters of the feeder pipes, including the velocity, wall shear stress, solution pH, temperature, concentration of dissolved iron, quality of solution, etc. For effective chemical decontamination of these thick oxides containing radionuclides such as Co-60, it is necessary to understand the corrosion behaviors of feeder pipes and the characteristics of oxide formed on it. In this work, we investigated the growth of oxide films that develop on type SA-107 Gr. B carbon steel in high temperature water and steam environment by scanning electron microscopy (SEM) and glow discharge optical emission spectrometry (GD-OES) for the quantification and the solidstate speciation of metal oxide films. This study was especially focused to set the experimental tests conditions how to increase the oxide thickness up to 50 m by changing the oxidation conditions, such as solution chemistry and thermo-hydraulic conditions both temperature and pressure and so on.

For the disposal of radioactive waste from nuclear facilities, assessing their radioactivity inventories is essential. As a result, countries with nuclear facilities are implementing assessment schemes tailored to their respective policies and available resources for radioactive waste management. This paper specifically describes the assessment scheme for radioactivity inventory applied to metal waste generated during the dismantling of the Japan Power Demonstration Reactor (JPDR), a 1.25 MW BWR. The distinctive aspect of the Japanese approach lies in the fact that, for a pair of a key nuclide and a difficult-to-measure (DTM) nuclide that lack a significant correlation in their concentrations, the mean activity concentration method was used. In this method, an arithmetic average of all measurements of the DTM nuclide from representative drums, including MDAs (Minimum Detectable Activities), was assigned to the concentration of the DTM nuclide for all drums, regardless of the concentration of its paired key nuclide. Conversely, for a specific pair of a key nuclide and a DTM nuclide with a significant correlation, the scaling factor method was applied, as is common in many other countries. This Japanese case can serve as a valuable reference for Korea, which does not have the option of using the mean activity concentration method in its assessment scheme.

Among the nuclear power plant facility improvement projects, out of a total 10 replacement reactor vessel closure head (RRVCH), five have been replaced, starting with Gori Unit 1, and five, including Hanul Unit 1, Hanbit Units 5 and 6, and Hanul Units 3 and 4 will be replaced in the future. This paper presents the method of treating Latch Housing among radioactive waste generated during the replacement of Hanul Unit 2 (February 2023). Latch Housing controls the control rod by receiving magnetic force from the CRDM’s Coil Stack. Located in the Old Reactor Vessel Head (ORVH) Hot Spot, the range of measured radiation dose rate was 0.3 to 0.8 mSv h-1 (up to 4.5 mSv h-1). The amount of radioactive waste generated was 35.8 Baler-Drum (based on 200L), and the order of treatment was to cut into the Omega Seal of the CRDM, the CRDM and Latch Housing Transfer, the boundary of the CRDM and Latch Housing, the Rod Travel Housing, the Motor Housing and the Latch Assembly, and then transfer and Drumming. In the United States, out of 93 operating reactors, 31 reactor vessel heads have been replaced and 19 reactor vessel heads are scheduled to be replaced. In Korea, 25 reactors are in operation, and two reactors have been permanently shut down. Among them, the nine old reactors for more than 30 years (as of September 2021) are expected to achieve ALARA and reduce radwaste management costs through the management method applied to replace the reactor vessel head.

Recently, as carbon-neutral energy sources become increasingly important worldwide, SMRs (Small Modular Reactors), which offer significantly enhanced safety, versatility, and mobility compared to conventional nuclear reactors, are gaining attention as a viable alternative. SMR generally refers to small modular reactors with a power output of 300 MWe or less. Unlike conventional reactors, SMRs are characterized by an all-in-one design where peripheral systems and equipment are all integrated into the reactor itself, leading to enhanced reliability and durability. Additionally, the nuclear fuel reloading cycle is significantly extended compared to traditional reactors, resulting in a substantial reduction in maintenance difficulty and costs. Researchers have taken note of these characteristics of SMRs, particularly the extended fuel reloading cycle. Therefore, we have initiated the initial design of an ultra-small Micro Modular Reactor with an electricity generation capacity of 10 MWe and a fuel cycle of up to 55 years, with the goal of using it as a propulsion power source for various transportation modes, especially ships. Our design of MMR, called ‘ARA,’ is primarily distinguished by its use of U233 and Th232 fuels instead of conventional UO2 fuel. Due to various features of ‘ARA,’ including different fuel compositions, ARA is predicted to exhibit several characteristic features compared to conventional PWRs. In this study, among these characteristics, we focused on predicting changes in material composition within the fuel rod during the extended cycle operation of high-enriched fuel, rather than short-cycle operation using low-enriched fuel, unlike conventional reactors. The primary goal of this research is to observe the behavior of the composition of the materials used in the fuel cycle of the MMR, which utilizes U233 and Th232 fuels instead of UO2. Considering the difficulties in the spent nuclear fuel disposal process, many different trials were made to minimize the fission products of ARA, which differs from conventional reactors in terms of fuel type, size, and fuel cycle, in relation to waste generation.

Hydride reorientation is widely known as one of the major degradation mechanisms in Zirconium cladding during dry storage. Some previous theoretical models for hydride reorientation used assumption of an ideal radial basal pole orientation for HCP structure of Zirconium cladding. Under this assumption, circumferential hydride was considered to precipitate in the basal plane while radial hydride was considered to precipitate in the prismatic plane, thereby giving energetical penalty on thermodynamical precipitation of radial hydrides. However, in reality, reactor-grade Zirconium cladding exhibits average 30° tilted texture, adding complexity to the hydride precipitation mechanism. In this study, reactor-grade Zirconium cladding was charged with hydrogen and hydride reorientation -treated specimens were fabricated. Microstructural characterization of hydrides was conducted via following three methods in terms of interface and stored energy. And this study aimed to compare these characteristics between circumferential and radial hydrides. Using Electron Back Scattered Diffraction (EBSD), the interface was investigated assuming that interface lies parallel to the axial axis of the tube. These were further validated with Transmission Electron Microscope (TEM). In addition, Differential Scanning Calorimetry (DSC) analysis was conducted to calculate the stored energy. This investigation is expected to establish fundamental understanding of how hydrides precipitate in Zirconium cladding with different orientations. And it will also increase the predictability of radial hydride formation and help understanding the mechanical behavior of Zirconium cladding with radial hydrides.

Over the past decades, particle physics has made significant progress in characterizing neutrinos even if neutrinos have extremely small cross-section (~10-44 cm2), allowing them to penetrate any object. More recently, neutrino detection and analysis have indeed become valuable tools in various aspects of nuclear science and technology. Neutrinos are detected using various methods, including Inverse Beta Decay (IBD), Neutrino-electron scattering, and Coherent Neutrino-Nucleus Scattering (CNNS). For the detection of anti-neutrinos from nuclear reactor, the Inverse Beta Decay (IBD) is commonly considered with scintillators. Notable experiments in Korea, such as RENO and NEOS, have been conducted using the IBD method at the Hanbit Nuclear Power Plant since 2006. Additionally, the NEON experiment, which employs CNNS, which has a significantly larger reaction cross-section than IBD but its low-energy signal detection difficulty, has been ongoing since 2021. Based on the results of NEOS (2015-2020) the signal to noise is ~30 and IBD detection rate is ~2000 counts per day. The IBD event in nuclear power plants provides valuable information about reactor behavior. IBD count rates are in good agreement with the thermal power of the reactor. Furthermore, the neutrino energy spectrum can be used to estimate the fission isotope ratio of the reactor core, showing promise for obtaining reactor core information from antineutrino detection techniques. Neutrino detection in nuclear facilities provides valuable information about reactor behavior. However, as a surveillance technology neutrino detection faces challenges due to the very low cross-section, requiring efforts to overcome limitations related to detector size and signal acquisition time. In 2008, the International Atomic Energy Agency (IAEA) included neutrino detection in its Research and Development (R&D) program for reactor safeguards. In January 2023, the IAEA organized a “Technical Meeting on Nuclear Data Needs for Antineutrino Spectra Applications” to discuss the latest developments and research results in this field. In summary, the use of neutrino detection in the nuclear field, particularly for reactor monitoring and safeguarding, has advanced significantly. Ongoing research and collaboration are expected to enhance our understanding of neutrinos and their applications in nuclear science and technology.

Korean innovative SMR has been implemented developing with improved safety/economy and i- SMR technology development project to secure a competitive edge in SMR. For nuclear power plants, according to the revision of the Nuclear Safety Act (2013.6), it is mandatory to be reflected in the aging management program of nuclear power plants, and the aging management and regulation of major nuclear power plants are being strengthened. For i-SMR, chemistry environment and management strategy is essential to mitigate corrosion and radiation fields, since it has compacted and integrated module designs. Since 1994, zinc injection into the reactor coolant system (RCS) has been applied more than 100 PWRs in the world to mitigate primary water stress corrosion cracking (PWSCC) and to reduce outof- core radiation fields. In domestic NPPs, 7 have been applying zinc injection and had up to 90% radiation field reductions. For this reason, SMR needs to apply zinc injection for chemistry strategy. Zinc target concentration will be 5~40 ppb at i-SMR, based on Ni-Fe-Cr materials as same as PWRs. Zinc injection location is in volume and purification control system between the volume control tank and charging P/P where the pressure is moderate. Zinc injection skid can consist of two micro-controllable pump (one for operation and one for stand-by) and one injection tank (batching tank for zinc solution). Zn, Ni, Si, Fe, and activated corrosion products should be monitored to identify zinc injection controls and trends. Flux mapping for core performance monitoring should be evaluated. The application of zinc will be essential and effective and bring sustainable reliability for corrosion control and mitigation strategy to meet the risk-free i-SMR development.

In the pressurized water nuclear reactors (PWRs), the upper and bottom head penetration nozzles, the geometric asymmetry of the welded part increases from the center to the outer part, increasing the possibility of defects. For this reason, it is important to perform early detection and management through analysis of defects occurring in the welded parts of upper and bottom penetration nozzles of reactor vessel. However, it is very difficult to operate boat sampling of the welding area because the spacing of the penetration nozzle of the bottom head of the reactor is very narrow. In addition, it is more difficult to collect welded specimens of bottom penetration nozzles by electrical discharge machining in a boric acid water environment of nuclear reactor. In this work, to overcoming these technical difficulties, we developed a boat sampling robot system, which is composed of the specimen collection electrode head, borate-mediated discharge electrode and control system. Also, we performed basic performance tests and summarize the results.

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

To evaluate the characteristics of radioactive waste from permanently shut down nuclear power plants for decommissioning, there is a method of directly analyzing samples and, on the other hand, a computerized evaluation method based on operation history. Even if the radioactivity of the structures or radioactive wastes in the nuclear power plant is analyzed by the computerized evaluation method, a method of directly analyzing the sample must be accompanied in order to more accurately know the characteristics of the nuclear power plant’s radioactive waste material. In order to obtain such samples, we need a way to collect materials from radioactive waste. However, in the case of a permanently shut down nuclear power plant with a long operating history, human access is limited due to radiation of the material. In this study, we propose a method of remotely collecting samples that guarantees radiation protection and worker safety at the site where radioactive waste is located.

In this study, in relation to the demolition of the building as a research reactor, in order to establish a basic design for preparation for relocation and installation of the TRIGA Mark-II, the present conditions such as actual measurements and structural safety were investigated, as well as technologies and cases related to the relocation and installation of cultural properties. Based on this, the basic design for the relocation and installation of cultural assets was established by reviewing the disassembly and transport design of the TRIGA Mark-II and the basic plan for the relocation site. Although the structural safety of the current self-weight of the structure is judged to be reasonable, when lifting the structure, it is necessary to consider a method of lifting the foundation by reinforcing the foundation so that the tensile force can be minimized in the structure. As for the technology to be applied before TRIGA Mark-II, the technology before non-transplacement was confirmed as the most reasonable method in terms of preserving the original form, securing safety, and securing economic feasibility. Among the non-replacement technologies, the methods that can be applied before reactor 1 can be largely classified into three types. The three methods to be reviewed can be largely classified into the traditional rail movement method, the movement method using transport equipment, and the crane movement method. Each required period was calculated from the basic design results, and the modular trailer method was judged to be the most efficient. From the basic design results, the required period for each stage according to the mobile construction method was calculated. Depending on the calculation result, the modular trailer method is judged to be the most efficient. However, the final construction method should be selected according to the detailed design results. Overall, the results obtained through this study suggest that it is possible to create a memorial hall without the previous installation of TRIGA Mark-II if the structure foundation is composed independently of the building foundation after conducting a detailed characteristic investigation on the foundation of the TRIGA Mark-II structure.

According to IAEA PRIS, there is no record of dismantling commercial heavy water reactors among 57 heavy water reactors around the world. In Canada, which has the largest number of heavy water reactors, three of the 22 commercial heavy water reactors with more than 500 MW are permanently suspended, Gentilly unit 2 (2012), Pickering unit 2 (2007), and Pickering unit 3 (2008), all of which chose a delayed decommissioning strategy. On the other hand, Wolsong unit 1, which will be the world’s first heavy water reactor to be dismantled commercially, will be immediately carried out as a decommissioning strategy. KHNP has established various cooperation systems with advanced companies and international organizations related to overseas NPP decommission and is actively exchanging technologies. Among them, the most important focus is on research cooperation related to COG (CANDU owners Group). The first case is a joint study on Conceptual Calandria Segmentation. Four areas of process, waste management, ALARA, and cost for decommissioning reactors to be submitted to Canadian regulators for approval of Pickering and Gentilly-2’s preliminary decommissioning plan have been evaluated, and research on Wolsong unit 1 is currently underway. The second case is Decommissioning and long-term waste management R&D. Although the technical maturity is low, it studies the common interests of member companies in the decommissioning of heavy water reactor power generation companies and long-term waste management. Robotics for dismantling high-radiation structures, C- 14, H-3 measurement and removal methods, and concrete decontamination technology, which are characterized by heavy water, are being actively studied. KHNP is strengthening international cooperation with COG to prepare for the successful decommissioning of Wolsong unit 1. Based on previous studies by Pickering and Gentilly-2, an evaluation of the decommissioning of Wolsong unit 1 reactor is being conducted. In addition, it is preparing for decommissioning through experience analysis of the pressure tube replacement project.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

As the decommissioning of nuclear power plants progresses, interest in the inevitably generated radioactive waste is also increasing. Especially, because the containers of ILW packages are significantly more expensive than the containers of LLW packages, the special attention should be focused on minimizing the number of the containers of ILW packages. The radiation dose limit for packaging of ILW shall not exceed 2 mSv/h and 0.1 mSv/h on contact and at 2 m, respectively in South Korea. Meanwhile, The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging and cutting plan, the number of containers for ILW packages generated from decommissioning of Reactor Vessel Internal (RVI) of a nuclear power plant that has been in operation for decades was optimized in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

A phosphorylation (phosphate precipitation) technology of metal chlorides is considering as a proper treatment method for recovering the fission products in a spent molten salt. In KAERI’s previous precipitation tests, the powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides containing actinides such as uranium and rare earths with lithium phosphate in a molten salt was known as solidliquid reaction. In order to increase the precipitation reaction rate the powder of lithium phosphate dispersed by stirring thoroughly in a molten salt. As one of the recovery methods of the metal phosphates precipitated on the bottom of the molten salt vessel cutting method at the lower part of the salt ingot is considered. On the other hand, a vacuum distillation method of all the molten salt containing the metal phosphates precipitates was proposed as another recovering method. In recent study, a new method for collecting the phosphorylation reaction products into a small recovering vessel was investigated resulting in some test data by using the lithium phosphate ingot in a molten salt containing uranium and three rare earth elements (Nd, Ce, and La). The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is very slower than that of test using lithium phosphate powder. In this presentation, the precipitation reactor design used for phosphorylation reaction shows that the amount of molten salt transferred to the distillation unit will reduce by collecting all of the metal phosphates that will be generated using lithium phosphate powder into a small recovering vessel.