Cement is widely used as representative industrial material. In Korea, about 50 million tons of cement are consumed every year. In the manufacture of cement, raw materials containing NORM such as fly ash and bauxite are used. Therefore, the workers can be subjected to radiation exposure. The major exposure pathway in NORM industries is internal exposure due to inhalation of aerosol. Internal radiation dose due to aerosol inhalation varies depending on physicochemical properties of the aerosol. Therefore, the objective of this study was to investigate aerosol properties influencing inhalation dose in cement industries. In this study, aerosol properties were measured for two cement manufacturers. A particulate size distribution and concentration at various processing areas in cement manufacturing industries in Korea were analyzed using a cascade impactor. The mass density of raw materials and byproducts were measured using pycnometer. Shape of particulates was analyzed using SEM. The radioactivity concentration of Ra-226, Ra-228 for U/Th decay series was measured using HPGe. Particulate concentration by size was distributed log-normally with maximum at particle size about 7.2 μm in manufacturer A and 5.2 μm in manufacturer B. The mass density of fly ash and cement were 2.3±0.06, 3.2±0.02 g/cm3 respectively in manufacturer A. In manufacturer B, the mass density of bauxite and cement were 3.4±0.02, 2.9±0.01 g/cm3 respectively. The shape of particulates appeared as spherical shape in manufacturer A and B regardless of sampling area. Thus, a shape factor of unity could be assumed. The radioactivity concentrations of Ra-226, Ra-228 were 82±9, 82±8 Bq/kg for fly ash, and 25±4, 23±3 Bq/kg for cement in manufacturer A. In manufacturer B, the radioactivity concentrations of Ra-226, Ra-228 were 344±34, 391±32 Bq/kg for bauxite, and 122±13, 145±12 Bq/kg for cement. The radioactivity concentrations of Ra-226, Ra-228 in cement were less than raw materials such as fly ash and bauxite. It is because the dilution of the radioactivity concentration occurred during mixing with other raw materials in cement production process. This study results will be used as database for accurate dose assessment due to airborne particulate inhalation by workers in cement industries.

In emergency situations such as nuclear accidents or terrorism, radioactive and nuclear materials can be released by some environmental reasons such as the atmosphere and underground water. To secure the safety of human beings and to respond appropriately emergency situation, it is required to designate high and low dose rate regions in the early stages by analyzing the location and radioactivity of sources through environmental radiation measurement. This research team has developed a small gamma probe which is featured by its geometrical accessibility and higher radiation sensitivity than other drone detectors. A plastic scintillator and Silicon Photomultiplier (SiPM) were applied to the probe to optimize the wireless measurement condition. SiPM has a higher gain (higher than 106) and lower operating voltage (less than 30 V) compared to a general photodiode. However, the electronic components in the SiPM are sensitively affected by temperature, which causes the performance degradation of the SiPM. As the SiPM temperature increases, the breakdown voltage (VBD) of the SiPM also increases, so the gain must be maintained by applying the appropriate VBD. Therefore, when the SiPM temperature increases while the VBD is fixed, the gain decreases. Thus, the signal does not exceed the threshold voltage (VTH) and the overall count is reduced. In general, the optimal gain is maintained by cooling the SiPM or through a temperature compensation circuit. However, in the developed system, the hardware correction method such as cooling or temperature compensation circuit cannot be applied. In this study, it was confirmed that the count decreased by up to 20% according to the increase in the temperature of the SiPM when the probe was operated at room temperature (26°C). We propose methods to calibrate the total count without cooling device or compensation circuit. After operating the probe at room temperature, the first measured count is set as the reference value, and the correction factor is derived using the tendency of the count to decrease as the temperature increases. In addition, since this probe is used for environmental radiation monitoring, periodic measurements are more suitable than continuous measurements. Therefore, the temperature of the probe can be maintained by adding a power saving interval to the operation sequence of the probe. These two methods use the operation sequence and measurement data, respectively. Thus, it is expected to be the most effective method for the current system where the temperature compensation through hardware is not possible.

When the nuclear accident like the Fukushima is occurred, it is required to immediately determine the location of radioactive materials and their activities. Various studies related the unmanned technique to detect and characterize the contaminated area have been conducted. The Korea Institute of Nuclear Nonproliferation and Control (KINAC) has developed a new gamma detection system which consists of nine probes using a silicon photomultiplier (SiPM) and plastic scintillator. The probe is the small gamma detector designed to be carried and dropped near the accident area by the unmanned aerial vehicle. In this paper, we developed the improved design related to the angular dependence of the radioactive contamination detection system with the purpose of increasing the detection efficiency. The detection efficiency, radiation shielding and back-scattering varies depending on the direction of incidence of radiation because the probe has vertical structure of consisting scintillator, photomultiplier, and electric circuits. That is, when the experimental conditions are same except the direction of gamma probe, the result of measurements is different. It causes errors in measuring the radioactivity and location of the radioactive source. Since the direction of the probe is arbitrarily determined during the deployment of the probe through the unmanned aerial vehicle, it is considered changing the design of the scintillator from a conventional 1.0" × 1.0" Φ cylindrical shape to a 1.0" Φ spherical shape. In case of using the spherical scintillator, it is confirmed that angular dependence was reduced through MCNP simulation. The difference in the measurement depending on the direction of the probe could be reduced through additional structure design. Finally, we hope that the developed detection system which has the probes with spherical shape of scintillator can measure the radioactivity and location of the radioactive source in a range of about 100 × 100 m2 by measuring for at least 5 minutes. The field test at Fukushima area will be carried out with JAEA members in order to prove the feasibility of the new system.

In liquid scintillation counting, sample radioactivity is analyzed by measuring photons emitted from counting vials. Quenching effect lowers photon intensity from samples, which leads to lower counting efficiency. So an appropriate quenching correction according to characteristics of samples is important. In this study, the quenching correction for H-3 analysis was conducted according to the characteristics of paper packaging material leached samples. The leached samples are made from H-3 leaching method which is in the process of development for H-3 contamination screening. There are several ways of quenching correction such as internal standard (IS) method, quench correction curve and triple-to-double coincidence ratio (TDCR) method, etc. For quench correction curve, quenched standard set, which has the same matrix as experimental samples, is needed to be prepared. Each leached sample, however, has different matrix and color depending on condition of leaching experiment, which means that it is not capable of preparing standard set having same matrix with the samples. In this study, the counting samples are used for plotting quench correction curve instead of quenched standard set. Spectral quench parameter of the external standard [SQP(E)] is used as quench indicating parameter (QIP). TDCR and counting efficiencies determined by IS method are used as counting efficiencies. The quench curve of TDCR versus SQP(E) has R2 = 0.55 and the curve of efficiency from IS method versus SQP(E) has R2 = 0.99. TDCR is known for approximate counting efficiency, however, TDCR as counting efficiency needs careful use for H-3 analysis of leached samples. The curve used efficiency from IS method is suitable for H-3 analysis of leached samples. In this study, the quench correction curve is prepared for H-3 analysis of leached samples of paper packaging material. SQP(E), TDCR and efficiency from IS method was used as parameters to plot the quench correction curve, and, the efficiency from IS method is suitable for H-3 analysis of the leached samples. The result of this study can be used for H-3 analysis of leached samples of paper packaging material.

In Korea, Kori Unit 1, a commercial pressurized water reactor (PWR), was permanently shut down in June 2017, and an immediate decommissioning strategy is underway. Therefore, it is essential to understand the characteristics of radioactive waste during the decommissioning process of nuclear power plants (NPP). Because radioactive waste must be handled with care, radioactive waste is treated in a hot cell facility. Hot cell facility handles radioactive waste, and worker safety is essential. In this study, it was dealt with whether or not the radiation safety regulations were satisfied when processing the core beltline metal of the dismantling waste treated at the post irradiation examination facility (PIEF) of the hot cell facility. Core beltline metal used for the pressure vessel in the reactor is carbon steel, and it is continuously irradiated by neutrons during the operation of the NPP. A radiological safety estimation of the behavior of radioactive aerosols during the cutting process within the PIEF was carried out to ensure the safety of the environment and workers. When processing the core beltline metal in PIEF, dominant six nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) of aerosol are generated. Accordingly each cutting device, amount of aerosol and value of dose is different. Using a 99.97% efficiency HEPA filter, the emission concentration of the dominant nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) in the air source term was satisfied with the emission control standard of Nuclear Safety Commission No. 2016-16. It was confirmed that the radioactivity concentration in the airborne source term inside the PIEF is in equilibrium state, when ventilation is considered. Also, the mass of aerosol and the concentration of airborne source term differed according to the thickness of the saw blade of the cutting tool, and the exposure dose of the worker was different through Monte Carlo N-Particle (MCNP). At that time, 60Co accounted for 95.4% of the exposure dose, showing that 60Co had the highest impact on workers, followed by 55Fe with 2.7%. The worker’s dose limit is satisfied in accordance with Article 2 of the Nuclear Safety Act and the dose limit of radiation-controlled area is found to be satisfied in accordance with Article 3 of the rules on technical standards for radiation safety management at this time.

Water electrolysis is an efficient method to enrich heavy hydrogen isotopes (tritium and deuterium) in the aqueous phase. Although an alkaline water electrolyzer has been commercialized for mass production of hydrogen, such a method requires additional purification steps to remove electrolytes from the final concentrates. On the other hand, proton exchange membrane water electrolysis (PEMWE) does not require additional electrolyte treatment steps, and PEMWE is operated at higher current density compared to the alkaline water electrolysis. In this study, we investigated deuterium and tritium separation from light water by PEMWE. Separation behaviors at the anode and cathode were analyzed, and H/D and H/T separation factors were compared.

Plasma Arc Melter (MSO) system has been developed for the treatment and the stabilization of various kinds of hazardous and radioactive waste into the readily disposable solidification products. Molten salt oxidation system has been developed for the for the treatment of halogen- and sulfurbearing hazardous and radioactive waste without emissions of PCDD/Fs and acid gases. However, PAM system has showed some difficulty in the off-gas treatment system due to the volatilization of radionuclides and toxic metals at extremely high-temperature plasma arc melter and the emissions of acid gases. MSO system has also showed the difficulty in the treatment of spent molten salt into the disposable waste form. Present study discussed the results of organics destruction performance tests for the PAM-MSO combination system, which is proposed and developed to compensate the drawbacks of each system. The worst-case condition tests for the organics destruction were conducted at lowest temperatures and the worst-case condition tests for the retention of metals and radionuclides were conducted at highested temperatures under the range of normal operating condition. For the worst-case organic destruction test, C6H5Cl was selected as a POHCs (Principal Organic Hazardous Constituents) because of its high incinerability ranking and the property of generation of chlorine gases and PCDD/Fs when incompletely destroyed. Simulated concrete waste spiked with 1 L of C6H5Cl was treated and the emissions of 17 kinds of PCDD/Fs and other hazardous gases such as CO, THCs, NOx, SO2 and HCl/Cl2 were measured. For the worst-case condition tests for the retention of metals and radionuclides, Pb and Cs were selected because of its high volatility characteristics. The emissions of PCDD/Fs was extremely lowered than the emission limit and those of other hazardous constituents were below their emission limit. The results of performance tests on the organics destruction suggested that tested PAM-MSO combination system could readily treat PCBs-bearing spent insulation liquid, spent ion-exchange resins used for the treatment of spent decontamination liquid in the decommission process and the concreted debris bearing hazardous organic coating materials. The decontamination factor of Cs and Co were 1.4×105, 1.4×105, respectively. The emisison of Pb was 0.562 ppm. These results suggested that tested PAM-MSO system treated low-level radioactive and pb-bearing mixed waste.

The segmentation of activated components including reactor vessel and reactor vessel internals requires many information. The primary information is material composition, trace materials in the composition, neutron flux during operation, etc. According to the EPRI report the primary basis of activity in a decommissioning source term is the activated metals from the reactor vessel and vessel internal components. The report indicates that over 95% of the radioactivity from decommissioning, except from spent nuclear fuel, consists of activated metals. These are from the reactor vessel, reactor internal structures and expendable components which are constructed primarily of various grades of stainless steel. Stainless steel contains appreciable levels of impurity cobalt. The common primary radionuclides of concern for the disposal environment from activated metals identified in US and international studies include C-14, Cl-36, Ni-59, Co-60, Ni-63, etc. The most common types of stainless steels used in reactor vessel construction and internal components include the Type 304(L), Type 316(L) and various grades of Inconel. The components of stainless steel are mainly Ni, Cr, Mo, Nb, etc., and when these elements are activated, they produce nuclides such as Nb-94, Tc-99, Sr-90, etc. In this study, the current status of activation analysis is reviewed to understand the effects of many variables. Also, the effect of trace materials is reviewed, including transformation of radioactive nuclides.

In this research, KPS manufactured Full System Decontamination (FSD) equipment, which is consisted of Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS), Chemical Waste Decomposition & Treatment System (CWDS) and conducted demonstration test to prepare Decontamination and Decommissioning (D&D) project of Kori nuclear power plant in Korea. Each equipment of FSD was modularized due to the limited size of equipment hatch of Kori nuclear power plant. To simulate the expected circumstances in nuclear power plant such as usage of heater or position of each equipment, additional equipment was used. The chemical concentration and flow rate of process water for FSD were used as mentioned in the previous study by KHNP CRI. FSD was conducted for three cycles and each cycle was consisted of oxidation, reduction, chemical decomposition and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of UV lamp and IX by the heat. Total volume of process water for FSD demonstration test was 2.5 m2. KPS conducted decontamination performance review by calculating thickness reduction and weight loss of installed specimen. Operational review was conducted as if FSD test was conducted in the field based on the result of demonstration test. One of the most prioritized features is the workers’ safety. Also, the appropriate position of equipment needs to be considered to meet the required specification of component.

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.

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.

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.

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.

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.

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.

A significant amount of piping is embedded in nuclear power plants (NPPs). In decommissioning these materials must be removed and cleaned. It can then be evaluated for radioactivity content below the release level. MARSSIM presents Derived Concentration Guideline Levels (DCGLs) that meet release guidelines. Calculating DCGL requires scenarios for the placement of embedded pipe and its long-term potential location or use. Some NPPs choose to keep the embedded pipes in the building. Because others will dismantle the building and dispose of the piping in-situ, determining the disposal option for embedded piping requires the use of measurement techniques with the sensitivity and accuracy necessary to measure the level of radioactive contamination of embedded piping and meet DCGL guidelines. The main measuring detectors used in NPPs are gas counters that are remotely controlled as they move along the inside of the pipe. The Geiger-Mueller (GM) detector is a detector commonly used in the nuclear field. Typically, this GM detector used 3-detectors that cover the entire perimeter of the pipe and are positioned at 120-degrees to each other. This is called a pipe crawler. It is very insensitive to gamma and X-ray, only measures beta-emitter and does not provide nuclide identification. The second method is a method using a high-resolution gamma-ray detector. Although not yet commercialized in many places, embedded piping is a scanning method. The technique only detects gamma-emitting nuclides, but some nuclides can be identified. Gamma-ray scanning identifies the average concentration per pipe length by the detector collimator. It is considerably longer than a pipe crawler. In addition, several techniques, including direct measurement of dose rate and radiochemical analysis after scraping sampling, are used and they must be used complementary to each other to determine the source term. Expensive sampling and radiochemical analysis can be reduced if these detectors are used to measure the radioactivity profile and to perform waste classification using scaling factor. In the actual Trojan NPP, a pipe crawler detector was used to survey the activity profile in a 26 foot of an embedded pipe. These results indicate that the geometric averaging of the factors and a dispersion values for each nuclide are constant within the accuracy factors. However, in order to accurately use the scaling factor in waste classification, it must have sample representativeness. Whether the sample through smear or scraping is representative of the radionuclide mixture in the pipe. Since the concentration varies according to the thickness of the deposit and depending on the location of the junction or bend, a lot of data are needed to confirm the reliability of the nuclide mixture. In this study, the reliability of the scaling factor, sampling representativeness and concentration measurement accuracy problems for waste classification in decommissioning NPP were evaluated and various techniques for measuring radioactive contamination on the inner surface of embedded pipes were surveyed and described. In addition, the advantages and limitations of detectors used to measure radioactivity concentrations in embedded piping are described. If this is used, it is expected that it will be helpful in determining the source term of the pipe embedded in the NPPs.

The decommissioning project of NPP is a large-scale project, with various risks. Successful implementation of the project requires appropriate identification and management of risks. IAEA considered risk management “To maximize opportunities and to minimize threats by providing a framework to control risk at all levels in the organization”. Framework-based risk management allows project managers to identify key areas in which action should be taken at an appropriate time. Also, it enables effective management of projects by supporting decision-making on sub-uncertainty. Risk could be categorized according to the source of the risk. This is called Risk Breakdown Structure (RBS), and is documented as a risk assumption register through a risk identification process. IAEA considers various factors when defining risks in accordance with ISO 31000:2009. IAEA SRS No.97 presents a recommended risk management methodology for the strategy and execution stage of the decommissioning project of nuclear facilities through the DRiMa project conducted from 2012 to 2015. The risk breakdown structure classified in DRiMa project is as follows: (1) Initial condition of facility, (2) End state of decommissioning project, (3) Management of waste and materials, (4) Organization and human resources, (5) Finance, (6) Interfaces with contractors and suppliers, (7) Strategy and technology, (8) Legal and regulatory framework, (9) Safety, and (10) Interested parties. They have various prompts for each category. Such a strategy for dealing with risks has negative risks (threats) or positive risks (opportunities). The negative risks are as shown in avoid, transfer, mitigate and accept. On the other side, the positive risks are as shown in exploit, share, enhance and accept. During the decommissioning, a contingency infrastructure is needed to decrease the probability of unexpected events caused by negative risks. The contingency infrastructure of decommissioning project includes organization, funding, planning, legislation & regulations, information, training, stakeholder involvement, and modifications to existing programs. Since all nuclear facilities have different environmental, physical or contamination conditions, risks and treatment strategies should also be applied differently. This risk management process is expected to proceed at the stage of establishing and implementing a detailed plan for the decommissioning project of each individual plant.

In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

Solid radioactive waste such as rubble, trimmed trees, contaminated soil, metal, concrete, used protective clothing, secondary waste, etc. are being generated due to the Fukushima nuclear power plant accident occurred on March 11, 2011. Solid radioactive waste inside of Fukushima NPP is estimated to be about 790,000 m3. The solid radioactive waste includes combustible rubble, trimmed trees, and used protective clothing, and is about 290,000 m3. These will be incinerated, reduced to about 20,000 m3 and stored in solid waste storage. The radioactive waste incinerator was completed in 2021. About 60,000 m3 of rubble containing metal and concrete with a surface dose rate of 1 mSv/h or higher will be stored without reduction treatment. Metal with a surface dose rate of 1 mSv/h or less are molten, and concrete undergoes a crushing process. About 60,000 m3 of contaminated soil (0.005 ~1 mSv/h) will be managed in solid waste storage without reduction treatment. The amount of secondary waste generated during the treatment of contaminated water is about 6,500 huge tanks, and additional research is being conducted on future treatment methods.