Detectors used for nuclear material safeguards activities are using scintillator detectors to quickly calculate the uranium enrichment at various nuclear material handling facilities. In order to measure the uranium enrichment, a region of interest is set around 185.7 keV which is the main gamma emission energy of uranium-235 in which the proportional relationship between the amount of uranium-235 and the net count is used. It is necessary to perform channel/energy calibration that a specific channel of the multi-channel analyzer is set to 185.7 keV. Most detector manufacturers have a built-in calibration source so that it is automatically performed when the detector starts to operate. In addition, the scintillator detector requires attention because the channel/energy gain may change depending on the ambient temperature so that a calibration source is used to compensate for this. In this paper, the spectral features are examined from among the scintillator detectors seeded with calibration sources used for safeguards activities. For this purpose, FLIR’s Identifinder-2 R400 T2 model and Canberra’s NAID model were used. HM-5 contains about 15nCi of Cs-137 and a photoelectric peak occurs at 662.1 keV. NAID contains about Am-241 of 55 nCi which alpha decays and subsequently emits gamma rays of 59.5 keV and 26.3 keV. The major difference among the detectors occurs in the background spectrum due to the difference in the source. From that kind of spectral features, it can be confirmed that the equipment is operating properly only when the spectrum by the corresponding calibration source is accurately known. The results of this study will enable a better understanding of the characteristics of scintillator detectors used for uranium enrichment analysis. Therefore, it is expected to be used as basic research for related software utilization as well as development in the future.

Since 2018, Central Research Institute of Korea Hydro & Nuclear Power (KHNP–CRI) has been operating an X-ray irradiation system with a maximum voltage of 160 kV and 320 kV X-ray tube to test personal dosimeters in accordance with ANSI N13.11-2009 “Personnel Dosimetry Performance- Criteria for Testing”. This standard requires that dosimeters for the photon category testing be irradiated with the X-ray beams appropriate to the ISO beam quality requirements. KHNP-CRI has implemented the fourteen X-ray reference radiation beams in compliance with ISO-4037-1, 2, and 3. When installing the X-ray irradiation system, KHNP-CRI evaluated the uncertainties of dose conversion coefficients for deep and shallow doses, based on “Catalogue of X-ray spectra and their characteristic data – ISO and DIN radiation qualities, therapy and diagnostic radiation qualities, unfiltered X-ray spectra” published by Physikalisch Technische Bundesanstalt (PTB). A CdTe detector (X-123, AMPTEK) with disk type collimators made of tungsten was used to acquire X-ray spectra. The detector was located at 1 m from the center of the target material in the Xray tubes. Six uncertainty factors for the dose conversion coefficients for the fourteen X-ray beams were chosen as follows; the minimum and maximum cut-off energies Emin and Emax, the air density (ρ), the accuracy of the high-voltage of the X-ray tube, statistics of the pulse height spectra and the unfolding method. For example, uncertainty of each quantity for a HK30 beam was calculated to be 0.3%, 2.32%, 0.19%, 1.25%, and 0.13%, and 0.18%, respectively. The combined standard uncertainty for the deep dose conversion coefficient of the HK30 beam was calculated to be 2.67%. The coverage factor corresponding to a 95 percent confidence interval was obtained as k = 1.8 using a Monte Carlo method, which is slightly lower the coverage factor of k = 1.95 for a Gaussian distribution. This seems to result from that two dominant uncertainties, the unfolding uncertainty and minimum cut-off energy uncertainty, follow a rectangular distribution.

Natural radionuclides-containing substances (NORM) contain natural radionuclides and cause radiation exposure. In Korea, safety management measures were needed to deal with and dispose of radon mattresses containing monazite in relation to such NORM. However, there is no clear safety management system related to NORM waste in Korea. In order to manage this reasonably and systematically, it is necessary to investigate and analyze standards and management measures related to the treatment and disposal of NORM waste. Therefore, this study investigated and analyzed the exemption and clearance level of NORM waste regulations in international organizations and foreign countries. IAEA GSR Part 3, 2013/59/Euratom, ANSI/HPS N13.53, CRCPD SSRCR Part N, and ARPANSA Publications 15 safety management regulations were analyzed to investigate safety management standards for NORM waste. The exemption and clearance level in international organizations and foreign countries were compared and analyzed based on radioactive concentration and dose. In addition, the management measures proposed for each literature were also investigated. As a result of the analysis, IAEA GSR Part 3 applied 1 mSv as a regulatory exemption level, 1 Bq/g for uranium and thorium series as a clearance level, and 10 Bq/g for K-40 nuclides. The IAEA recommends a differential approach to the potential and scale of exposure. The EU applied 1 Bq/g to uranium and thorium families and 10 Bq/g to K-40 nuclides for both regulatory exemption and clearance levels. The EU recommended that it be managed in proportion to the scale and likelihood of exposure as a result of the action. It is analyzed that this is similar to the IAEA’s management plan. In the United States, there was no single federal government radioactive concentration and dose for NORM management. The management plan differed in management status and level from state to state, and K-40 was excluded from regulation unless it was intentionally enriched. In the case of Australia, the radioactive concentration of uranium and thorium was 1 Bq/g as a standard for regulatory exemption and 1 mSv as a dose. As a management plan, it was suggested to dispose of waste by means of accumulation, dilution/dispersion, and reclamation. It was also suggested that the scale of exposure, like international organizations, take into account the possibility. The results of this study are believed to be used as basic data for presenting domestic NORM waste treatment and disposal methods in the future.

As nuclear power plants are operated in Korea, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated. Due to the increase in the amount of radioactive waste generated, the demand for transportation of radioactive wastes in Korea is increasing. This can have radiological effect for public and worker, risk assessment for radioactive waste transportation should be preceded. Especially, if the radionuclides release in the ocean because of ship sinking accident, it can cause internal exposure by ingestion of aquatic foods. Thus, it is necessary to analyze process of internal exposure due to ingestion. The object of this study is to analyze internal exposure by ingestion of aquatic foods. In this study, we analyzed the process and the evaluation methodology of internal exposure caused by aquatic foods ingestion in MARINRAD, a risk assessment code for marine transport sinking accidents developed by the Sandia National Laboratory (SNL). To calculate the ingestion internal exposure dose, the ingestion concentrations of radionuclides caused by the food chain are calculated first. For this purpose, MARINRAD divide the food chain into three stages; prey, primary predator, and secondary predator. Marine species in each food chain are not specific but general to accommodate a wide variety of global consumer groups. The ingestion concentrations of radionuclides are expressed as an ingestion concentration factors. In the case of prey, the ingestion concentration factors apply the value derived from biological experiments. The predator's ingestion concentration factors are calculated by considering factors such as fraction of nuclide absorbed in gut, ingestion rate, etc. When calculating the ingestion internal exposure dose, the previously calculated ingestion concentration factor, consumption of aquatic food, and dose conversion factor for ingestion are considered. MARINRAD assume that humans consume all marine species presented in the food chain. Marine species consumption is assumed approximate and conservative values for generality. In the internal exposure evaluation by aquatic foods ingestion in this study, the ingestion concetration factor considering the food chain, the fraction of nuclide absorbed in predator’s gut, ingestion rate of predator, etc. were considered as influencing factors. In order to evaluate the risk of maritime transportation reflecting domestic characteristics, factors such as domestic food chains and ingestion rate should be considered. The result of this study can be used as basis for risk assessment for maritime transportation in Korea.

At Nuclear Power Plant (NPP), aging management is performed as part of the Periodic Safety Review (PSR) in accordance with the Nuclear Safety Act. The purpose of the aging management program (AMP) is to manage the integrity of structures, systems and components (SSCs) in NPPs over time and use. Through this, aging deterioration is mitigated to increase equipment life and secure long-term operation safety. Fuel Oil Chemistry is one of the AMPs. Through this program, aging management is performed for storage tanks, piping and other metal components that contact with diesel fuel oil. The program is focused on managing loss of material due to general, pitting, crevice, and microbiologically-influenced corrosion (MIC) and fouling that leads to corrosion of the diesel fuel tank internal surfaces. The fuel oil aging management method currently applied to NPPs in Korea measures the concentration of water and particulate contamination in the oil, analyzed the trend, and periodically cleans and inspect the inside of tanks. Among them, in monitoring MIC, a direct analysis and monitoring of the amount of microorganisms may be more effective. In this study, a method for improving the MIC monitoring system for diesel fuel oil systems was reviewed by reviewing reference documents including NUREG 1801 and examining the methods actually applied in US NPPs.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.

Around the world, Nuclear Power Plants (NPPs) have been operated since the 1950s and are used as a major power source. In Korea, Kori unit 1 stared commercial operation for the first time in 1978, and as of 2023, 25 units of NPPs are in operation. NPPs produce electricity for about 40 to 60 years after receiving an operating license, and after securing safety through a safety evaluation, the operating period is extended. NPPs that operate for a long time are systematically evaluated for safety at regular intervals through Periodic Safety Review (PSR) recommended by the IAEA. In Korea, PSR has been introduced and performed since 2000. This study reviewed the process of the PSR by comparing with the international PSR procedure. The PSR process is established through the IAEA SSG-25 document and proceeds in the order of establishment of basis document - individual factor evaluation - global assessment - integrated improvement plan. In Korea, PSR is carried out in a similar process, but there are some differences from the IAEA’s procedure. The safety factor review is conducted under the agreement of basis document between the licensee and the regulatory body, but the prior agreement procedure with the regulatory body is not reflected in Korea. As a result, if the licensee and the regulatory body have different opinions on the current licensing basis and the modern safety standards after the evaluation is performed, a difference may occur in the review results and safety enhancement items, which may lead to inefficient PSR progress. PSR is conducted for the continuous safe operation and management of NPPs, and it is important to refer to overseas standards and cases. Although procedures, guidelines, and regulatory requirements are in place in Korea, continuous review and improvement are required. It is necessary to improve procedures such as basis document and global assessment in order to more efficiently carry out PSR evaluation by regulatory agency and licensee’s safety enhancement actions of domestic NPPs

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

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

Our research team has developed a gamma ray detector which can be distributed over large area through air transport. Multiple detectors (9 devices per 1 set) are distributed to measure environmental radiation, and information such as the activity and location of the radiation source can be inferred using the measured data. Generally, radiation is usually measured by pointing the detector towards the radioactive sources for efficient measurement. However, the detector developed in this study is placed on the ground by dropping from the drone. Thus, it does not always face toward the radiation source. Also, since it is a remote measurement system, the user cannot know the angle information between the source and detector. Without the angle information, it is impossible to correct the measured value. The most problematic feature is when the backside of the detector (opposite of the scintillator) faces the radiation source. It was confirmed that the measurement value decreased by approximately 50% when the backside of the detector was facing towards the radiation source. To calibrate the measured value, we need the information that can indicate which part of the detector (front, side, back) faces the source. Therefore, in this study, we installed a small gamma sensor on the backside of the detector to find the direction of the detector. Since this sensor has different measurement specifications from the main sensor in terms of the area, type, efficiency and measurement method, the measured values between the two sensors are different. Therefore, we only extract approximate direction using the variation in the measured value ratio of the two sensors. In this study, to verify the applicability of the detector structure and measurement method, the ratio of measured values that change according to the direction of the source was investigated through MCNP simulation. The radioactive source was Cs-137, and the simulation was performed while moving in a semicircular shape with 15 degree steps from 0 degree to 180 degrees at a distance of 20 cm from the center point of the main sensor. Since the MCNP result indicates the probability of generating a pulse for one photon, this value was calculated based on 88.6 μCi to obtain an actual count. Through the ratio of the count values of the two sensors, it was determined whether the radioactive source was located in the front, side, or back of the probe.

The increasing use of drones in terrorist attacks highlights the need for effective strategies to prevent and respond to drone terrorism. This study uses machine learning approach to identify factors that predict the success of drone terrorism and suggests policy alternatives for preventing such acts. Drone terrorism is becoming increasingly accessible due to advancements in information and communication technology, and events such as North Korea’s drone infiltration and the Russia-Ukraine war demonstrate the potential threat of drone attacks on Important National Facilities, including nuclear power plants. Using the Global Terrorism Database (GTD), this study analyzed drone terrorism incidents that occurred worldwide from 2016 to 2020. The study employed the Random Forest algorithm, which can incorporate multiple factors and their interactions, making it particularly suitable for social science research. The study provides new insights by deriving predictors that were previously overlooked in empirical analyses of drone terrorism. The findings of this study can aid in the establishment of anti-terrorism policies aimed at addressing the growing threat of drone terrorism. This can include the organization and expansion of the crisis management governance terrorism response council, the creation of a working manual through the partial revision of laws concerning drone terrorism response, and the implementation of anti-drone equipment and systems. Ultimately, the insights gained from this study can provide development of effective strategies aimed at preventing and responding to drone attacks. The study highlights the importance of proactive measures to mitigate the risks posed by drone technology in the context of terrorism.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

During decommissioning and site remediation of nuclear power plant, large amount of wastes (including radioactive waste) with various type will be generated within very short time. Among those wastes, soil and concrete wastes is known to account for more than 70% of total waste generated. So, efficient management of these wastes is very essential for effective NPP decommissioning. Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes from the generation. The system is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. By adopting modular type, the system is good for dealing with variable situation where system capacity needs to be expanded or contracted depending on the decommissioning schedule, good for minimizing secondary waste generated during maintenance of failed part and also good for disassemble, transfer and assemble. The contamination evaluation module of the system has two sub module. One is for quick measurement with NaI(Tl) detector and the other is for accurate measurement with HPGe detector. For waste transfer, the system adopts LTS (Linear Transfer System) conveyor system showing low vibration and noise during operation. This will be helpful for minimizing scattering of dust from the waste container. And for real time positioning of waste container, wireless tag was adopted. The tag also used for information management of waste history from the generation. Once a container with about 100 kg of soil or concrete is loaded, it is moved to the weight measurement module and then it transfers to quick measurement module. When measured value for radioactivity concentration of Co- 60 and Cs-137 is more than 1.0 Bq/g, then the container is classified as waste for disposal and directly transferred to stacking and storage rack. Otherwise, the container is transferred to accurate measurement module. At the accurate module, the container is classified as waste for disposal or waste for regulatory clearance depending on the measurement result of 0.1 Bq/g. As the storage rack has a sections for disposal and regulatory clearance respectively, the classified containers will be positioned at one of the sections depending on the results from the contamination evaluation module. The system can control the movement of lots of container at the same time. So, the system will be helpful for the effective nuclear power plant decommissioning in view of time and budget.

In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

Disposal of radioactive waste requires radiological characterization. Carbon-14 (C-14) is a volatile radionuclide with a long half-life, and it is one of the important radionuclides in a radioactive waste management. For the accurate liquid scintillation counter (LSC) analysis of a pure beta-emitting C-14, it should be separated from other beta emitters after extracted from the radioactive wastes since the LSC spectrum signals from C-14 overlaps with those from other beta-emitting nuclides in the extracted solutions. There have been three representative separation methods for the analysis of volatile C-14 such as acid digestion, wet oxidation, and pyrolysis. Each method has its own pros and cons. For example, the acid digestion method is easily accessible, but it involves the use of strong acids and generates large amount of secondary wastes. Moreover, it requires additional time-consuming purification steps and the skillful operators. In this study, more efficient and environment-friendly C-14 analysis method was suggested by adopting the photochemical reactions via in-situ decomposition using UV light source. As an initial step for the demonstration of the feasibility of the proposed method, instead of using radioactive C-14 standards, non-radioactive inorganic and organic standards were investigated to evaluate the recovery of carbon as a preliminary study. These standards were oxidized with chemical oxidants such as H2O2 or K2S2O8 under UV irradiations, and the generated CO2 was collected in Carbo-Sorb E solution. Recovery yield of carbon was measured based on the gravimetric method. As an advanced oxidation process, our photocatalytic oxidation will be promising as a time-saving method with less secondary wastes for the quantitative C-14 analysis in low-level radioactive wastes.

When decommissioning a nuclear power plant, a large amount of radioactive waste is generated simultaneously. Therefore, efficient treatment of radioactive waste is crucial to the success of the decommissioning process. An utility or decommissioning contractor of NPP often build separate radioactive waste treatment facilities (RWTF) to handle this waste. In Korea, RWTFs are planned to be built for the decommissioning of the Kori Unit 1 and Wolsong Unit 1. In this study, we introduce an application case of using process simulation to derive the optimal layout design and investment plan for a radioactive waste treatment facility. In particular, the steam generator is the largest and most complex device processed in RWTF. Therefore, it is necessary to reflect the large equipment processing area that can treat steam generators in the design of RWTF. In this study, Siemens’ Plant Simulation® was used to derive an optimization plan for the dismantling area of large equipment in RWTF. First, a virtual facility was built by modeling based on the steam generator dismantling process and facilities developed by Doosan Enerbility. This was used to pre-validate the facility investment plan, discover wasteful factors in the logistics waste streams, and evaluate alternatives to derive, validate, and apply appropriate improvement alternatives. Through this, we designed a layout based on the optimal logistics waste streams, appropriate workstations, and the number of buffer places. In addition, we propose various optimization measures such as investment optimization based on optimal operation of facility resources such as facilities and manpower, and establishment of work standards.

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

The nuclear power plant decommissioning project inevitably considers time, cost, safety, document, etc. as major management areas according to the PMBOK technique. Among them, document management, like all projects, will be an area that must be systematically managed for the purpose of information delivery and record maintenance. In Korea, where there is no experience in the decommissioning project yet, data management is systematically managed and maintained during construction and operation. However, if the decommissioning project is to be launched soon, it is necessary to prepare in consideration of the system in operation, what difference will occur from it in terms of data management, and how it should be managed. As a document that can occur in the decommissioning project, this study was considered from the perspective of the licensee. Therefore, the types of documents that can be considered at Level 1 can be divided into (1) corresponding documents, (2) project documents, (3) internal documents, and (4) reference materials. Four document types are recommended based on Level 1 for the classification of documents to be managed in the decommissioning of nuclear facilities. In this study, documents to be managed in the decommissioning project of nuclear facilities were reviewed and the type was to be derived. Although it was preliminary, it was largely classified into major categories 1, middle categories 2, and 3 levels, and documents that could occur in each field were proposed. As a result, it could be largely classified into corresponding documents, project documents, internal documents, and reference materials, and subsequent classifications could be derived. Documents that may occur in the decommissioning project must be managed by distinguishing between types to reduce the time for duplication or search, and the capacity of the storage can be efficiently managed. Therefore, it is hoped that the document types considered in this study will be used as reference materials for the decommissioning project and develop into a more systematic structure.