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.

In the field of 3H decontamination technology, the number of patent applications worldwide has been steadily increasing since 2012 after the Fukushima nuclear accident. In particular, Japan has a relatively large number of intellectual property rights in the field of 3H processing technology, and it seems to have entered a mature stage in which the growth rate of patent applications is slightly reduced. In Japan, tritium is being decontaminated through the Semi-Pilot-class complex process (ROSATOM, Russia) using vacuum distillation and hydrogen isotope exchange reaction, and the Combined Electrolysis Catalytic Exchange (CECE, Kurion, U.S.) process. However, it is not enough to handle the increasing number of HTOs every year, so the decision to release them to the sea has been made. Another commercial technology in foreign countries is the vapor phase catalyst exchange process (VPCE) in operation at the Darlington Nuclear Power Plant in Canada. This process is a case of applying tritium exchange technology using a catalyst in a high-temperature vapor state. The only commercially available tritium removal technology in Korea is the Wolseong Nuclear Power Plant’s Removal Facility (TRF). However, TRF is a process for removing HTO from D2O of pure water, so it is suitable only for heavy water with high tritium concentration, and is not suitable for seawater caused by Fukushima nuclear power plant’s serious accident, and surface water and groundwater contaminated by environmental outflow of tritium. Until now, such as low-temperature decompression distillation method, water-hydrogen isotope exchange method, gas hydrate method, acid and alkali treatment method, adsorption method using inorganic adsorbent (zeolite, activated carbon), separator method using electrolysis, ion exchange adsorption method using ion exchange resin, etc. have been studied as leading technologies for tritium decontamination. However, any single technology alone has problems such as energy efficiency and processing capacity in processing tritium, and needs to be supplemented. Therefore, in this study, four core technologies with potential for development were selected to select the elemental technology field of pilot facilities for treating tritium, and specialized research teams from four universities are conducting technology development. It was verified that, although each process has different operating conditions, tritium removal performance is up to 60% in the multi-stage zeolite membrane process, 30% in the metal oxide & electrochemical treatment process, 43% in the process using hydrophilic inorganic adsorbent, and 8% in the process using functional ion exchange resin. After that, in order to fuse with the pretreatment process technology for treating various water quality tritium contaminated water conducted in previous studies, the hybrid composite process was designed by reflecting the characteristics of each technology. The first goal is to create a Pilot hybrid tritium removal facility with 70% tritium removal efficiency and a flow rate of 10 L/hr, and eventually develop a 100 L/hr flow tritium removal system with 80% tritium removal efficiency through performance improvement and scale-up. It is also considering technology for the postprocessing process in the future.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

In general, dose assessment must be performed to obtain approval for clearance of radioactive waste. If the annual dose criteria through dose evaluation satisfies the clearance condition, radioactive waste can be disposed of. Various programs are used to perform dose assessment. NRCDOSE GASPAR is used as a program to assess the amount of radiation exposed to atmospheric emissions. Program is easy to use and results can be checked immediately after execution. GASPAR requires main input factors by exposure route such as site specifics, source term, special location, block data. Basically, program has default input values but user can easily modify it. The most important factor is that when entering a nuclide, the effect on progeny radionuclides is not automatically calculated. User should consider the dose contribution from progeny radionuclides. In this study, dose assessment was performed for combustible waste incineration using NRCDOSE GASPAR. And it was confirmed that exposure dose of individuals and groups criteria for clearance regulation.

In KAERI, Waste storage facility in the radiation management area has stored a large amount of wood waste. The amount of waste is approximately 27,000 kg, it accounts for 17% of the total waste in waste storage facility. Proper disposal of wood waste improves the fire resistance performance, secure storage space and reduce disposal costs. In order to self-disposal of wood waste, it is necessary to satisfy the self-disposal standards stipulated by the domestic Atomic Energy Act and the treatment standards of the Waste Management Act. The main factors of standards are surface contaminant, radionuclide activity and radiation dose effects. To confirm the contamination of wood waste, direct indirect measurement methods and gamma nuclide analysis were performed. To evaluate radiation dose, various computational programs were used. The results of the analysis were satisfied with domestic regulations on the classification and self-disposal of radioactive wastes. Based on this results, KAERI submitted the report on wood waste self-disposal plan to obtain approval. After final approval, wood waste is to be incinerated and incineration ash is to be buried in the designated place. The objective of this study is to provide total procedure of wood waste self-disposal and effective representative sampling method.

There are various types of level gauging method such as using float, differential pressure, hypersonic, displacement and so on. In this study, among them, the method utilizing the differential pressure was reviewed. The strengths include: the differential pressure type level gauge can measure the level without direct contact of the sensor with media. That is to say, the level can be measured even if the sensor is far away from the tank. And regardless of the size of the tank, the level can be measured if the pneumatic pipes are installed. The weaknesses include: the sensor needs intermedium to recognize the level. The intermedium utilizes a fluid, which is compressed air. It is difficult to handle that compressed air has the properties of a gas. And to make compressed air needs compressor, tank and pneumatic pipes. So if you have many tanks, you need to install exponentially the pneumatic pipes. As well, level measurement range is limited to the points where the pneumatic pipes of the tank is installed. And if a compressed air that supplies to the sensor leaks, uncertainty will increase. A compressed air is colorless and odorless, so it’s difficult to pinpoint the leak. Finally, events like cracks and clogging can cause inaccurate measurement. Rather than using only differential pressure, it is better to use another measurement method according to the situation of the facility.

Radioactive waste is classified into Intermediate level, low level, and very low potential based on the amount of radioactivity per unit gram, that is, the concentration limit. This method of classifying radioactivity per unit weight is not a problem if all packaged wastes are homogeneous. However, the reality is that not all waste is homogeneous. Relative hotspots may exist. Also, when several items are mixed, if one item has a relatively higher concentration than other items, it can become a relative hotspot. In Korea, even if all nuclides in a single radioactive waste package satisfy the low level concentration limit, if even one nuclide exceeds the concentration limit, the radioactive waste package becomes the intermediate level. In case of the United States, the US NRC provides regulations related to obtaining license as well as presents the technical position on the average waste concentration called Concentration Averaging and Encapsulation Branch Technical Position (CA BTP). CA BTP classifies waste into four types : Blendable Waste, Encapsulated items, Single Discrete Items, and Mixture of Discrete Items, and presents each approach to concentration averaging. In general, this is a method that suggests an acceptable ratio in case of the waste, which relatively high concentration waste is mixed. In order to apply this in Korea, we compare the classification standards for low and Intermediatelevel waste in Korea and the United States, related laws and backgrounds, and the application methods of CA BTP.

Following a radioactive waste criterion and clearance level radioactive waste Act Article 2. “The radioactive wastes confirmed by the Commission as having concentration by nuclide not exceeding the value determined by the Commission through incineration, reclamation, recycling, etc”. The combustible clearance level radioactive wastes like lumbers are incinerated and non-combustible wastes like concreted are buried. The metals clearance level radioactive wastes are recycled after being re-molded. However, the clearance level radioactive waste with keeping its original forms is not common. Due to the nature of KAERI, the equipment are brought into the radiation-controlled zone for experiments. Those equipment are conservatively considered contaminated and categorized with radioactive waste following nuclear safety acts. In this case, the spectroscopy device which is clearance level radioactive waste is self-disposed for use in non-controlled areas. The 4 devices are composed of 3 gamma-ray spectroscopy and 1 alpha, beta counting system. Those devices were used for clearance level radioactive waste’s radioisotope analysis in Radioactive Waste Form Test Facility which is used in a separated room for analysis. This room will be released in nonradiation controlled area, therefore those devices will be moved to non-controlled area and keep using. Last April self-disposal was reported to the regulatory body and got acceptance last May. Those devices were moved to non-controlled area last July. This case will be good example for reuse equipment which stop using in radiation controlled area but can keep used.

The number of dismantled nuclear facilities is increasing globally. Dismantling of nuclear facilities generates large amount of waste such as concrete, soil, and metal. Concrete waste accounts for 70% of the total amount of waste. Since hundreds of thousansds of tons of concrete waste generated, securing technology of reduction and recycling of waste is emerging as a very important issue. The objective of this study is to synthesize geopolymer using inorganic materials from cement fine powder in concrete waste. Dismantled concrete waste contains a large amount of calcium silicate hydrate(C-S-H), Ca(OH)2, SiO2, etc., which is an inorganic material required for the synthesis of geopolymer. SiO2 affects the compressive strength of the geopolymer and Ca(OH)2 affects the curing rate. A high concentration of alkali solution is used as an alkali activator, and alkali activator is necessary for the polymerzation reaction of metakaolinite. The experiment consists of three steps. The first step is to react with concrete waste and hydrochloric acid to extract ions. In the solid after filtration, SiO2 and Al2O3 are composed of 84.10%. It can be used instead of commercial SiO2 required for the synthesis of geopolymer. The second step is to add NaOH up to pH 10, impurities can be removed to extract Ca(OH)2 with high purity. The final step is to add NaOH up to pH 13, and Ca(OH)2 extraction. The alkali solution generated after the last reaction can be recycled into an alkali activator during the synthesis of the geopolymer. If dismantled concrete waste is recycled during geopolymer synthesized, the volume reduction rate of dismantled concrete waste is more than 50%. If you put the radioactive waste in the recycled solidification materials synthesis from concrete waste by dismantling of nuclear facilities, it is possible to reduce the amount of waste generated and disposal costs.

In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.