Trojan Nuclear Power Plant (NPP), a four-loop PWR designed by Westinghouse and owned by Portland General Electric (PGE), reached its initial threshold in 1975 and was operational until November 1992. PGE received a Possession Only License from the NRC in May 1993. In 1995, limited decommissioning activities began at the Trojan, including the completion of a large components removal project to remove and dispose of four steam generators and pressurizers from the containment building. In April 1996, the NRC approved a plan to dismantling the Trojan NPP and began more aggressive component removal activities. At the end of 1998, part of the radioactive drainage system began to be removed, and embedded piping decontamination and survey activities began. Trojan NPP has more than 8,840 m of contaminated pipelines throughout the power block. Most of Trojan NPP’s contaminated embedded piping can generally be divided into four categories drainage piping, ventilation ducts, buried process piping, and other items. For the Trojan NPP, the complete removal of contaminated and embedded piping without damaging the building would have significantly increased costs due to the structural considerations of the building and the depth of the embedded pipe. Therefore, Trojan NPP has chosen to conduct the Embedded Pipe Remediation Project (EPRP) to clean and in situ survey of most of the embedded piping to meet the Final Site Survey (FSS) acceptance criteria, with much success. This study provides a discussion of EPRP activities in the Trojan NPP, including classification and characterization of affected piping, modeling of proposed contamination acceptance criteria, and evaluation of various decontamination and survey techniques. It describes the decontamination tools, techniques, and survey equipment and the condition of work and cost estimate costs used in these projects. To identify embedded piping and drains at the Trojan NPP, based on frequent site surveys, plan sketches showing an overview of system flow paths and connections and database were developed to identify drain inputs and headers. This approach effort has been a successful method of remediation and site survey activities. The developed database was a valuable asset to the EPRP and a Work Breakdown Structure (WBS) code was assigned to each drains and headers, allowing the embedded piping to be integrated into the decommissioning cost estimation software (Decon. Expert) and schedule, which aided in decommissioning cost estimation. Also, regular database updates made it easy to check the status of the decommissioning project data. The waste system drain at Trojan NPP was heavily contaminated. The goal of the remediation effort is to completely remove all removable contamination and to reduce the fixed contamination below the decided contamination acceptance criteria. Accordingly, Hydrolysis, Media blast, Chemical decontamination and Pipe removal were considered as remediation option. Trojan NPP’s drainage pipe decontamination option did not cause a significant corrosion layer inside the pipe and media blast was chosen as the main method for stainless steel pipe. In particular, the decommissioning owner decontaminates most of the embedded piping in-situ to meet the FSS acceptance criteria for economic feasibility in Trojan NPP. The remaining pipe was filled with grout to prevent leaching and spreading of contamination inside the pipe. In-situ decontamination and survey of most of these contaminated pipes are considered the most cost-effective option.

The radioactive Sr-90, which is formed from beta decay, is well known as one of the most commonly detected nuclides in radioactive waste. In 2015, it was reported that Sr-90 was observed in some soil and metal wastes among the 516 drums of radioactive waste transferred from the decommissioning site of the Korea Research Reactor (in Seoul) to the disposal site (in Gyeongju). Decontamination and sequestration of radionuclides, including Sr, from nuclear waste is important because they are hazardous and harmful to the ecological environment. Immobilization of these nuclides using a zeolite framework is suitable and simple method that has been widely studied. Therefore, it is still necessary to continuously explore the thermal stability of various zeolites and environmental changes around adsorbed cations in zeolite pore for effective immobilization of these radionuclides. In this study, we observed the thermal stability in fully Sr-exchanged natrolite (Sr-NAT), one of small-pore zeolite, from room temperature to 350°C using the in-situ synchrotron X-ray powder diffraction and thermogravimetric (TGA) analysis. In addition, we investigated the structural changes in Sr-NAT during temperature increase by Rietveld analysis. Sr-NAT exhibited apparent zero thermal expansions (ZTE) with the thermal expansion coefficients of -3(1) × 10-6 at the initial stage of increasing the temperature due to dehydration process. In the section from 250°C to 300°C, a phenomenon like negative thermal expansion (NTE) occurs in which the unit cell volume of Sr-NAT decreases despite the increase in temperature. Sr-NAT maintained well its crystallinity up to 350°C, and it became amorphous at 350°C. In this study, we provide a fundamental understanding of the structural changes and thermal stability mechanism of Sr-exchaged zeolite natrolite with increasing temperature.

Radiological characterization is important in decommissioning and dismantling of nuclear facilities, in order to assess the radioactivity concentration, classify the wastes, and secure workers’ safety. The Some components such as Reactor Pressure Vessel (RPV) in nuclear facilities has dose rate higher than Sv/hr, thus in-situ gamma spectroscopy systems suffer from a very high count rate which causes energy resolution degradation, photo-peak shift, and count loss by pile-up and dead-time. The system must be operated in a very high count rate, in order to measure spectra precisely and to quantify radionuclide contents. In order to apply in-situ measurement in high radiation dose rate environment, the sensor, front-end electronics, and data acquisition (DAQ) should be carefully selected and designed as well as precise design of collimators and radiation shield. In this paper, the components of the detector system were selected and performance was evaluated in a high count rate before design the collimator and shield. A LaBr3 coupled with a PMT having short decay time constant (16 nsec) was selected for high count rate application, and two different amplifiers (a conventional charge sensitive preamplifier with 50 usec decay time constant, and wide-band voltage amplifier) were tested. As DAQs, DT5781 (14 bit, 100 MS/s, CAEN) of Pulse Height Analysis (PHA) which is conventionally used signal processing method in the gamma spectroscopy, and DT5730 (14 bit, 500MS/s, CAEN) of Pulse Shape Discrimination (PSD) which is similar to Charge to Digital Convertor (QDC) were used. The number of photons incident to the detector was varied by changing the detector-source distance with Certificate Radiation Material (CRM), and compared to the output count rate. The count rate capability, and energy resolution with different amplifier and DAQ was evaluated. Additionally, the performance of DAQs in extremely high count rate was evaluated with signal data generated by the emulator which can simulate the detector signal waveforms fed into the DAQ based on the measured spectrum.

Reliable evaluation of radioactivity inventory for the nuclear power plant components and residual materials is very important for decontamination and decommissioning. This can make it possible to define optimum dismantling approaches, to determine radioactive waste management strategies, and to estimate the project costs reasonably. To calculate radioactivity of the nuclear power plant structure, various information such as interest nuclide, cross-section, decay constant, irradiation time, neutron flux, and so on is required. Especially irradiation time and neutron flux level are very changeable due to cycle specific fuel loading pattern, the plant overhaul, cycle length. However most of the radioactivity calculations have generally been performed assuming one representative or average neutron flux during the lifetime of the nuclear power plant. This assumption may include excessive conservatism because the radioactivity level has the characteristics of saturation and decay. Therefore, considering these variables as realistically as possible could prevent overestimation. In order to perform realistic radioactivity calculation, we developed monthly relative power contribution factor applying plant-specific operation history and cycle-specific neutron flux. The factors were applied to the radioactivity calculation. The calculation results ware compared with measured values of the neutron monitors that were actually installed and withdrawn from the nuclear power plant. As a result of the comparisons, there are good agreements between the calculated values and measured values. These accurate calculation results of radioactivity could contribute to the establishment of radioactive waste dismantling strategies, the classification of radioactive waste, and the deposit of disposal costs for safe and reasonable decommissioning of the nuclear power plant.

The Korean administration assumed that the amount of low and medium level waste generated during the decommissioning of nuclear facilities in Korea was 14,500 drums (based on 200 L) and designed the LILW repository accordingly. Accordingly, it is necessary to separate the nuclear power plant decommissioning waste into clearance waste by mobilizing means such as decontamination and cutting as much as possible, and to deregulate it together with non-radioactive waste. As a result, clearance waste and non-radioactive waste are dominated by concrete and metal, and it is necessary to evaluate how to recycle them. Many existing studies have conducted research on each recycling method, and accordingly, it can be judged that the technological maturity is sufficient. Accordingly, we would like to propose a method for comprehensive management and evaluation of concrete. By applying the decision matrix proposed in IAEA TRS No. 401, it will be possible to compare the 5 factors (cost, technical feasibility, risk, availability of disposal, and full cycle impact). However, in the case of concrete, if the existing construction waste recycling methodologies are fully used, the technical feasibility can be considered equal. Therefore, it was judged that it would be good to introduce the aspect of public acceptance as an evaluation item instead of technical feasibility. The amount of waste that can be generated when decommission a nuclear power plant is only insignificant compared to the total amount of waste concrete that is generated during the year. Accordingly, one option is to fully integrate the waste concrete recycling system and utilize it for road construction. Next, it is possible to suggest the option of recycling in the construction of shields in the nuclear industry, as suggested in previous studies, and the method of using it as a backfill material such as for a decommissioned NPP site or other sites. As an example, and a draft stage, this study was evaluated based on existing studies after all options were equally weighted. When the profit and loss was evaluated in a way that a maximum of 5 points were given to each option, the case of using it as a backfill in various applications was evaluated as the best option. Unlimited recycling, such as road construction, was evaluated to be highly damaging in terms of public acceptance.

The goal of the decommissioning of nuclear facilities is to remove the regulations from the Nuclear Safety Act. The media that can be considered at the time of remediation stage may usually include soils, buildings, and underground materials. In addition, underground materials may largely be the groundwater, buried pipes, and concrete structures. In fact, it can be seen that calculations of the Derived Concentration Guideline Level (DCGL) and ALARA action levels was conducted in the case of overseas decommissioning experiences of Nuclear Power Plants (NPPs). Therefore, the aim of this study is to review the remediation activities and scenarios applied for the calculation of ALARA action level from the overseas decommissioned nuclear power plants. Media that can be considered for DCGL calculation at the time of license termination may differ from site to site. If the DCGL for the target media was derived, whether additional remediation actions are required under the DCGL value from the ALARA perspective was identified by calculating the ALARA action levels in the case of the U.S. The activities to determine whether additional clean-up is justified under the regulatory criteria are remediation actions which is dependent on the material contaminated. Therefore, the typical materials that can be subjected to remediation are soils and structure basements in the overseas cases. Remediation actions involved in the decommissioning process on the structure surfaces can be typically considered to be scabbling, shaving, needle guns, chipping, sponge and abrasive blasting, pressure washing, washing and wiping, grit blasting, and removal of contaminated concrete. For the cost-benefit analysis of the media subject to DCGL calculation, it is necessary to assume a scenario for the remediation actions of the target media. The scenarios can be largely divided into two types. Those are basement fill and building occupancy scenario. In basement fill mode, buildings and structures on the site are removed, and the effect of receptors from the contamination of the remaining structures is considered. In the building occupancy mode, it is assumed that the standing building remains on the site after the remediation stage. It is a situation to evaluate how the effect of additional remediation actions changes as the receptors occupy inside of the contaminated building. Therefore, parameters such as population density, area being evaluated, monetary discount rate, numbers of years, etc. can be set and assessed according to the scenarios.

Kori Unit 1, Korea’s first commercial nuclear power plant is preparing to dismantle after 40 years of power supply. However, unlike the public dose assessment for nuclear power plants in operation, the dose assessment for the public due to abnormal events during the decommissioning of nuclear power plants is insufficient. Therefore, in this study, the steam generator chamber is selected as hypothetical events target among metal waste, which is a major radioactive material generated during the decommissioning of nuclear power plant. In addition, the possible abnormal event scenarios and effective does to public in the Exclusion Area Boundary due to the released radioactive materials are predicted during the disassembly and transportation of the steam generator. For the source term that can be released during the dismantling of the steam generator, the inventory of each radionuclide is evaluated based on the smear test results of the steam generator replaced in Kori Unit 1 in 1998. To evaluate the diffusion of radioactive material, the atmospheric dispersion factor (χ/Q, sec/m3) is calculated through the PAVAN code of the US NRC using the meteorological data of the Kori nuclear power plant for 3 years from 2019 to 2021 according to IAEA recommendations. For the assessment of the public dose, the external dose coefficient and inhalation coefficient of the ICRP and the inhalation rate of the NRC Regulatory Guide 1.3 are referred. It is confirmed that the effective dose to the public in the Exclusion Area Boundary due to the abnormal event during the dismantling of the steam generator is much lower than the effective dose standard value of 250 mSv for 2 hours after the event in the Exclusion Area Boundary.

A simulation model was developed for heavy water pre-enrichment and detritiation by the Combined Electrolysis and Catalytic Exchange (CECE) process. In the CECE process, heavy water enrichment and detritiation are based on the principle that concentrated in to water phase through an isotopic exchange reaction between water vapor and hydrogen gas produced by a water electrolysis. An operational analysis for a liquid phase catalytic exchange column was carried out by the model equations, composed of a material balance and combined equilibrium relationships for a scrubbing and catalyst bed, respectively. As a result of simulation, the optimum flow ratio of water to the rising hydrogen gas in contact with the down-coming water was predicted as the key variables in the separation performance analysis at a given feed flow rate and isotopic composition. From a graphical approach based on this model, the operating conditions can be determined within the range where the operating line does not meet the combined equilibrium line before reaching the specified target concentration.

In preparation for the decommissioning of Kori unit 1 of the nuclear power plant (NPP), new containers of package, transportation, and disposal are being developed that reflect the type, generation amount, and radiological characteristics of decommissioning waste. The containers under development have internal volumes of 1 m3 ~ 14 m3 and loading weights of 1 ton ~ 35 tons, which are larger in size and have a higher loadable weight compared to the 200 L drum and IP-2 type transport container currently being used for packaging and transporting waste. So, there is a limit to handling new containers using existing transport systems (cranes, spreaders, forklifts, transport vehicles, etc.). Therefore, in this study, the status of handling equipment in NPP and disposal facilities was reviewed, the flow from the generation to disposal of decommissioning waste was analyzed, and the possibility of handling new container or the necessity of introducing new systems were derived. Except for some high-dose/high-radioactive wastes among decommissioning wastes, all wastes are finally disposed of through decommissioning area, temporary storage facility, waste treatment facility, waste storage facility, and receipt and storage building. The decommissioning area, temporary storage facility, and waste treatment facility are newly established areas for the decommissioning and should be equipped with a spreader/crane with a lifting weight of 15 tons, 35 tons, and 40 tons in consideration of the weight of the package to be handled in the zone. The waste storage facility has a 7.5 tons crane, so it can handle only some of the lower weight of the 5 to 35 tons package that is expected to be handled. Therefore, additional installation of spreaders/cranes, each with a lifting capacity of 15 tons and 40 tons, is required. The maximum loading weight of forklifts and transport vehicles operating at NPP, and disposal facilities is 10 tons and 12.6 tons, respectively. To transport the package, the facility must additionally install 15 tons and 40 tons forklifts, and 40 tons transport vehicles. Since the lifting weight of the crane installed on the transport vessel is also low at 12.5 tons, it is necessary to change the design of the existing or replace it with 40 tons to handle high-weight package. The results of this study will be used as basic data for the establishment of transport systems in the relevant area and facility, and design requirements for each equipment will be derived through additional research.

Reactor pressure vessels and steam generators generated in the process of dismantling nuclear power plants or replaced steam have various shape and occupy a considerable amount of the disposal site when disposed of in original shape. For the development of domestic technologies related to the disposal of large wastes, it is necessary to secure technologies for reducing large radioactive metal wastes, including technologies such as decontamination, cutting, melting, and residual radioactivity evaluation. Cases of disposal of steam generators in Europe and the United States were investigated. Except for u-tubes, steam generators are less contaminated or easily decontaminated, so it is possible to reduce the volume of waste subject to final disposal by exempting a significant amount through appropriate treatment. Korea Hydro & Nuclear Power Co. is currently temporarily storing 24 steam generators at 41.6 billion won. This paper presents a method to exempt more parts of the steam generator and reduce the volume of waste by properly combining mechanical cutting thermal cutting and melting to dispose of the steam generator. Currently the decontamination and dismantling industries of nuclear facilities are gradually expanding around the world. Therefore, it is necessary to localize the treatment technology for metal waste generated during maintenance and dismantling. The result of this study can be used to establish waste reduction and disposal method for dismantling steam generators.

Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.

Kori unit 1 was permanently shut down in 2007 and is currently awaiting approval for decommissioning and dismantling (D&D). The wastes generated during decommissioning is estimated to be approximately 14,500 of 200 L drums. In this study, the treatment process of decommissioning wastes will be reviewed through the case of the US Zion nuclear power station (ZNPS). Zion unit 1 and 2 received an operating license in 1973 and were permanently shut down and the spent nuclear fuel was transferred to the pool in 1998. The decommissioning was carried out according to the following five steps; (1) safe storage (SAFSTOR) dormancy, (2) preparation for decommissioning, (3) establishment of independent spent fuel storage installation (ISFSI) and transfer of the spent fuel and greater than class C radioactive materials, (4) decommissioning operations and (5) site restoration. The total volume of waste generated during decommissioning was expected to be approximately 1.7×105 m3. This is far above the Kori unit 1 waste estimation because ZNPS has a history of accidents and includes soil waste. Wastes were treated differently according to their properties and locations.

In NPP (nuclear power plant), boric acid is used as a neutron absorbent. So radioactive boric acid waste are generated from various waste streams such as discharge or leakage of reactor coolant water, floor drains, drainage of equipment for operation or maintenance, reactor letdown flows and etc. Depending on KHNP, 20,015 drum (200 L drum) of concentrated boric acid waste were stored in KOREA NPP until 2019. In previous study, our group suggested the waste up-cycling process synthesizing B4C neutron absorber using boric acid waste and activated carbon waste to innovatively reduce radioactive wastes. Radioactive activated carbon waste was utilized in off gas treatment system of NPP to capture nuclide such as I-131, C-14 and H-3. Activated carbon waste is treated as low-level radioactive waste and pre-treatment system for removing nuclide from the activated carbon waste is needed to use B4C up-cycling process. In this study, microwave treatment system is suggested to treat the activated carbon waste. Activated carbon waste was exposed to microwave for a few minutes and temperature of the waste was dramatically increased over 400°C. Nuclide in the activated carbon waste were selectively removed from the waste without massive production of secondary off gas waste.

Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

Hydrogen isotopes (H, D, T) separation technologies have received great interest for treatments of tritiated liquid waste produced in Fukushima. In addition, the separated deuterium and tritium can be utilized in various industries such as semiconductors and nuclear fusion as expensive and rare resources. However, separating hydrogen isotopes in gas and liquid forms still requires energyintensive processes. To improve efficiency and performance of hydrogen isotope separation, we are developing water electrolysis, cryosorption, distillation, isotope exchange, and hydrophobic catalyst technologies. Furthermore, an analytical method is studied to evaluate the separation of hydrogen isotopes. This presentation introduces the current status of hydrogen isotope research in this research group.

The decommissioning of Kori Unit 1 is expected to generate a large amount of clearance waste. Disposing of a large amount of clearance waste is economically costly, so a recycling method has emerged. However, clearance waste recycling is expected to cause many conflicts among various stakeholders. In the previous study, possible conflicts were selected in consideration of the domestic environment and major issues. Based on this, this study classifies stakeholders involved in conflicts by group, and suggests ways to enhance understanding by stakeholder and enhance reliability. In this study, stakeholders are classified into four groups that share the same conflicts, and each of the following measures is suggested. 1) Stakeholder Engagement. 2) Common understanding of radiation risks, dialogue between the public/recycling industry/ regulatory agency. 3) Incentives to promote recycling clearance waste. 4) Reliable outlet store for recyclable clearance waste. The above understanding enhancement measures are presented so that a solution to conflict can be smoothly derived when designing a clearance waste-related consultative body composed of interested parties in the future. As a more specific solution, measures to enhance stakeholder trust can be suggested for each understanding enhancement measure. Reliability enhancement measures are also presented so that they can be applied to each stakeholder group, and these are as follows. 1) Write a stakeholder engagement plan, Measures for stakeholder participation in measuring the radioactivity concentration of clearance waste. 2) Active use of easy-to-understand radioactivity comparison data, Expansion of information on environmental radiation dose to public, nuclear/radiation education, Held a tour event at the nuclear power plant decommissioning site, New website for clearance waste information disclosure. 3) Incentives for recycling industries in which the Ministry of Environment or KHNP partially bears the losses that occur when the sales rate is low. Incentives are provided to consumers by including recyclables of clearance waste for Green Card’s green consumption points. 4) Online outlets open for recyclable clearance waste with easy-to-understand radioactivity comparison data. It is expected that if the above-mentioned reliability enhancement measures are used, it will be possible to secure the trust of stakeholders and reduce the gap between stakeholders in the future clearance-related consultative body.

The RPV internal structure is a high radio activated part and has very complex geometry. Therefore, it needs to be cut remotely with an automated cutting system to minimize the worker exposures. To do so, we made up the remote laser cutting system with a laser cutter, robot manipulator and control software system and the laser cutter is moved by the robot manipulator based on the command from the control software system. A laser cutter is required to keep the desired standoff position between the nozzle of the laser cutter and surface of the cut target model to cut properly. Moreover, in the remote cutting process, an exact time and sequence control of the air supply and the laser emission is required for the cutting quality and the process safety. In this study, we proposed the PERT chart-based process execution and control methodology. The PERT chart is a graph which is represented by nodes and edges. The node of the PERT chart has the information about the activity details such as activity type, execution time and related device. Using the edge we make the sequence of the desired activity execution. A PERT chart of the cutting scenario is compiled in the control software system to creates data and thread structure to operate the physical device. We built software architecture to interpret and execute the PERT chart efficiently in the digital simulation platform which enables us to use existing pre-built simulation scenario for the laser cutting process. In addition, we have tested various laser cutting test cases in our test bed to verify the performance of our system. The test bed environment has the shape of the RPV internal structure and is placed under water.

Radioactive contamination of soil on the site of a nuclear facility has a characteristic that radioactive nuclides are adsorbed into the pores between soil particles, making it quite difficult to decontaminate. For this reason, research on the development of various decontamination processes is being actively conducted. In this study, among various decontamination studies, a soil decontamination process using supercritical carbon dioxide was presented. The decontamination process uses supercritical carbon dioxide as the main solvent, which has a higher penetration power than other materials. Therefore, the process consists of the process of desorbing and extracting the target radionuclides between particles of soil. However, since nuclides exist as ions in the soil, polar chelating ligand material was introduced as an additive to nonpolar supercritical carbon dioxide for smooth chemical reactions in the soil. Thereafter, from the viewpoint of improving process continuity and efficiency, an alcohol material was introduced as an auxiliary solvent for liquefaction of chelating ligand in a solid state. Through prior research on the selection of a solvent for liquefaction of chelating ligand, ethanol and 2-propanol were finally selected based on whether the chelating ligand was dissolved. However, if the auxiliary solvent in which the chelating ligand is dissolved is to be combined with radionuclides in the soil, it must first be well dissolved in supercritical carbon dioxide, the main solvent. Therefore, in this study, the solubility of ethanol and 2-propanol in supercritical carbon dioxide was measured and the suitability was evaluated. The temperature conditions were carried out at 40°C, the same as the previously designed decontamination process, and the measurement was conducted by adjusting the pressure and volume through a syringe pump and a variable volume device. In addition, solubility was measured based on the observation of the ‘cloud point’ in which the image becomes cloudy and then bright. As a result of the experiment, several solubility points were measured at a pressure of 150 bar or less. If the flow rate ratio of supercritical carbon dioxide and auxiliary solvent derived from the results is applied to the soil decontamination process, it is expected that the process efficiency will increase in the future.

In the pilot scale test, the two scale-up factors (Electric energy per order EEO, Electric energy per mass EEM) were conducted to design the Chemical Waste Decomposition & Treatment System (CWDS). The CWDS consist of two kind UV lamp reactors to improve the decomposition rate of oxalic acid, which are low pressure amalgam UV lamp and medium pressure UV lamp. The two reactors were connected in series, and the hydrogen peroxide is mixed through a line mixer at the front of the reactor and injected into the reactors. The CWDS was connected with the full system decontamination equipment to purify the residual oxalic acid after chemical decontamination process. The full system decontamination equipment were included Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS) to operate the Oxidation/Reduction decontamination process and purify the process water. After decontamination process, the waste water will be cooled down into the 40°C and passed through the UV reactor at 110 gpm with hydrogen peroxide injection. The concentration of waste water is expected oxalic acid 1,700 ~ 2,000 ppm, Iron 5 ~ 20 ppm. As a result of the CBD test in the laboratory with simulated waste liquid, the amount of Low pressure amalgam lamp UV dose required to decompose 95% of oxalic acid in 2 m2 waste water was up to 1,800 mJ/cm2. The amount of medium pressure lamp UV dose was up to 450 mJ/cm2 at the same condition. We conducted demonstration test using 2 m2 waste water after the oxidation/reduction decontamination process, the decomposition rate 95% was obtained by low pressure amalgam UV lamp and medium pressure UV lamp reactor each.

To transport radioactive waste generated during the decommissioning of Kori Unit 1, transport containers of various sizes are being developed. Since these radioactive decommissioning waste transport containers are larger than the specifications of the existing IP-2 type transport containers, which are for operational radioactive waste, design of the CHEONG-JEONG-NURI needs to be changed when transporting them to disposal facility using the CHEONG-JEONG-NURI, which carries operational radioactive waste. In this study, design changes of the CHEONG-JEONG-NURI, cargo hold modification plan for efficient loading of radioactive decommissioning waste transport containers and radioactive decommissioning waste container loading arrangement (plan) were evaluated during the design life period (year 2034). First, as only the IP-2 type transport container with a weight of 7.5 tons and size of 1.6 m (W) × 3.4 m (L) × 1.2m (H) can be loaded in the cargo hold, if only the decommissioning radioactive waste containers are to be loaded and transported, cargo hold needs to be reinforced. Second, when both the radioactive decommissioning waste transport container of the same size as the current operating radioactive waste transport container, and the radioactive decommissioning waste transport container of the same size as the ISO-type transport container are to be loaded in the cargo hold of the CHEONG-JEONG-NURI and transported, the overall design changes (cargo hold size and load reinforcement) are required. Third, since the safe working load of the CHEONG-JEONG-NURI crane is 12.5-tons, it shall be replaced with a ship crane of 35-tons or more to handle the decommissioning radioactive waste container smoothly, or a gantry crane used in general port facilities shall be installed. When replacing with a ship crane of 35-tons or more, ship buoyancy, ship stability, and ship structural safety shall be considered. The possibility of moving in all 4 directions for smooth operation, and the possibility of lifting the transport container to a position higher than the height of the CHEONG -JEONG-NURI shall be considered. Loading and transporting all decommissioning radioactive waste containers, which are the same size as IP-2 and ISO-type transport containers, in the cargo hold of the CHEONG-JEONG-NURI is uneconomical due to the need for overall design changes (cargo size and load reinforcement, etc.). Also, delay in delivery of the operation wastes is expected due to a long-term design change period. Therefore, it is considered reasonable to load and transport only the decommissioning radioactive waste transport container, which is the same size as the IP-2 transport container, in the cargo hold.