The Wolsong unit 1 decommissioning project is the world’s first commercial pressurized heavy water reactor decommissioning project. Although there is a lot of accumulated experience and technology for decommissioning of pressurized water reactors around the world, it can be said that there is great difficulty as there is lack of prior experience and reference materials for pressurized heavy water reactor. On the other hand, if the world’s first pressurized heavy water reactor project is completed, it is possible to enter the overseas market for pressurized heavy water reactor decommissioning. It is also a good opportunity to do so. Accordingly, the current status of operation, plans, and construction of infrastructure related to decommissioning of pressurized heavy water reactors in Canada, which can be said to be the home country of pressurized heavy water reactor, were reviewed. So, this study seeks to present considerations for entering the pressurized heavy water reactors decommissioning market in the future.

Kori unit 1 and Wolsong unit 1 were permanently shut down in 2017 and 2019, respectively. Both plants were decided to demolish the building without reuse. Large structures must be demolished after removing systems and components in the building, and in the case of large structures, thorough planning is required because of the large scale of work. Therefore, in this study, important considerations in the phase of the demolition plan of large structures when decommissioning were analyzed. The demolition of large structures at nuclear facilities is major one phase of work within a broader decommissioning plan. Furthermore, the actual demolition of the structure (i.e., physical process) represents the last step in a process that begins with extensive planning and analysis. The National Demolition Association (NDA) has provided checklist items that should be considered before the start of a commercial demolition project and/or in the bid process. Important Considerations in the Phase of the demolition plan of large structures when decommissioning of nuclear facilities are Site knowledge and programs, Engineering survey/demolition plan, Hazardous and radioactive materials, Open air demolition, Financial and project management, Permits, Code adherence, and Special programs, Disposal pathway, Final site condition. The results of this study can be used as a basis for the Planning large structures demolition of the Kori unit 1 and Wolsong unit 1.

The Derived Concentration Guideline Level (DCGL) is required to release the facility from the nuclear safety act at the stage of site restoration of the decommissioning nuclear power plant. In order to evaluate DCGL, there are various requirements, and among them, the selection of input parameters based on the application scenario is the main task. Especially, it is important to select input parameters that reflect site characteristics, and at this time, a single deterministic value or a probabilistic distribution can be applied. If it is inappropriate to apply a particular single value, it may be reasonable to apply various distributions, and the RESRAD code provides for evaluation using probabilistic methods. Therefore, this study aims to analyze the difference between the application of the deterministic method and the application of the probabilistic method to the area and thickness of the contaminated zone among the site characteristics data. This study analyzed the thickness and area of the contaminated zone, and in the case of thickness, the deterministic method was applied by changing the thickness at regular intervals from the minimum depth considered by MARSSIM to the thickness of the unsaturated zone identified in previous research data. In addition, a probabilistic analysis was performed by applying a distribution to the thickness of contaminated zone. Second, for the area of the contaminated zone, the dose was evaluated for each area in consideration of the areas to be considered when deriving Area Factor (AF), and the resulting change in DCGL was observed. As a result, the DCGL tends to decrease as the thickness increases, and it seems to be saturated when the thickness exceeds a certain thickness. Therefore, It was confirmed that the level of saturated values is similar to that of entering a probabilistic distribution, and in the case of a parameter that is reasonable to enter as a distribution rather than as a single value, it is sufficiently conservative to perform a probabilistic evaluation. In the case of area change, the DCGL evaluation result showed that the DCGL increased as the scale decreased. The magnitude of the change varies depending on the characteristics of each radionuclide, and in the case of radionuclides where external exposure gamma rays have a major exposure effect, the change is relatively small. It can be seen that the change in DCGL according to the area has the same tendency as the AF applicable to the survey unit for small survey units applied in the final status survey.

The radiation field generated in the primary cooling system of a nuclear power plant tends to increase in intensity as radionuclides bind to the oxide film on the internal surface of the primary system, which is operated at high temperature and pressure, and as the number of years of operation increases. Therefore, decontamination of the primary cooling system to reduce worker exposure and prevent the spread of contamination during maintenance and decommissioning of nuclear power plants uses the principle of simultaneous elution of radionuclides when the corrosion oxide film dissolves. In general, a multi-stage chemical decontamination process is applied, taking into account the spinel structure of the corrosion oxide film formed on the surface of the primary cooling system, i.e. an oxidative decontamination step is applied first, followed by a reductive decontamination step, which is repeated several times to reach the desired decontamination goal. Currently, permanganic acid is commonly used in oxidative decontamination processes to remove Cr from corrosion oxide films. In the reductive decontamination step to remove iron and nickel, organic acids such as oxalic acid are commonly used. However, organic acids are not suitable for the final radioactive waste form. A number of multi-stage chemical decontamination technologies for primary cooling systems have been developed and commercialized, including NP-CITROX, AP/NP-CANDECON, CANDERM, AP/NP-LOMI and HP/CORD-UV. Among these, HP/CORDUV is currently the most actively applied primary cooling system chemical desalination process in the world. In this study, KAERI has developed a new chemical decontamination technology that does not contain organic chemical decontamination agents, with a focus on securing an original technology for reducing the amount of decontamination waste while having equivalent or better decontamination performance than overseas commercial technologies, and compared it with the inorganic chemical agent-based HyBRID (Hydrazine Based Reductive Metal Ion Decontamination) chemical decontamination technology.

The periodic safety review (PSR), for all operating nuclear power plants in Korea, has been conducted in accordance with SSG-25, a guideline suggested by the IAEA, The PSR is performed through the review of the regulatory body after the operator’s self-evaluation. In order to guarantee a high level of safety in consideration of the changed environment, such as operating experience (OE) and technology development, it should be comprehensively and integratedly performed, and it is also carried out every 10 years after the operation permit. However, in case that all or part of the reactor facilities have been permanently shut down, such as Kori Unit 1 and Wolsong Unit 1, Around a half of reactor facilities are not in operation. The periodic safety evaluation may not be conducted for unused parts if there is no safety hazard and if there are some difficulties for applying periodic safety evaluation. In considering that the biggest purpose of PSR safety (by PSR definition of KINS guideline) is to improve and accumulated factors such as aging deterioration, facility change, operation experience, and technological development for operating nuclear power plants. It refers to a comprehensive safety evaluation that is periodically performed during the period of operation of a nuclear power plant. It is necessary to review whether PSR should be performed for a nuclear power plant that is permanently shut down after nuclear power plant operation is terminated. Also, in IAEA SSR 2/2 Rev1, it is defined that PSR is performed during the nuclear power plant operation period. “Requirement 12: Periodic safety review, Systematic safety assessments of the plant, in accordance with the regulatory requirements, shall be performed by the operating organization throughout the plant’s operating lifetime, with due account taken of operating experience and significant new safety related information from all relevant sources”. Recently, Kori Unit 1 and Wolsong Unit 1 were decided to permanently shut down in June 2017 and December 2019, and are currently being prepared for decommissioning. According to the Wolsong decommissioning plan, decontamination and demolition will be completed by 2032. The PSR for permanent shutdown of Kori Unit 1 was submitted to the regulatory body in December 2018 and is under approval review. In the case of the permanent shutdown PSR of Wolsong Unit 1, the project will be launched in May 2023 and the PSR will be submitted to the regulatory body in May 2024. In the case of Wolsong Unit 1, it is necessary to operate the various systems, including the systems related to the spent fuel storage tank, even during the period of permanent shutdown. Such as the heavy water related systems used in common with Wolsong Unit 2, are essential operating systems. Based on Basic Subject Index (BSI), 112 out of 218 systems require operation, indicating that about 50% of systems require operation even after permanent shutdown. Decommissioning of systems and equipment will begin after the transfer to modular air-cooled canister storage (MACSTOR) by the end of 2025, and then in-depth discussions will be needed whether PSR evaluation is meaningful.

In order to evaluate the integrity of the reactor pressure vessel, various test specimens necessary to identify irradiation embrittlement. The degree of irradiation embrittlement of the vessel material by neutrons, from the construction to the end of the life of the plant, is evaluated by a monitoring plan that called surveillance program (a series of all plans to analyze and evaluate embrittlement through various tests and analyzes by placing a test piece inside the reactor pressure vessel and taking out a piece at an appropriate time according to the number of operation years and taking necessary measures for safe operation). The reactor monitoring specimens for Kori Unit-1 are located by axis at S (57°), T (67°), R (77°), N (237°), P (247°) and V (257°). Six surveillance capsules are attached to the inside of the pressure vessel around the core and to the outside of the thermal shield. This surveillance container determines the withdrawal timing of the surveillance container according to the provisions of ASTM E185-82. In the monitoring test piece, there are neutron dosimeter materials to measure and evaluate the irradiated neutron flux, and Ni, Cu, Fe, Co-Al, Cd, and shielded Co-Al monitors are wired in the monitoring container. Each axial position is contained in a spacer hole. The neutron dosimetry monitor measures the neutron dose using isotopes produced by neutrons during operation of the reactor. The Al-Co specimen, which can evaluate the degree of radioactivity of cobalt, is located on the lower part of the specimen. The content of Co in the Al-Co specimen is 0.15%, and when expressed in ppm, it is 1,500 ppm, which is similar to the cobalt content of 1,414 ppm in the internal structure of the reactor vessel presented in NUREG-3474. If the radiation value of the Al-Co sample in the reactor monitoring specimen can be measured, the radiation value of the internal structure of the reactor can be indirectly compared. Since the monitoring specimen is located outside of the thermal shield, radiation should be less than that of the thermal shield. Korea Reactor Monitoring Technology performed gamma measurement on Al-Co specimens in 6 monitoring specimens, and although there are differences depending on the sample, it shows radioactivity values around the order of 1E+07 dps/g, or Bq/g. In conclusion, it is thought that using this measurement values, it is possible to verify the evaluation of internal structure radiation for Kori unit-1 decommissioning.

The Korea Research Institute of Standards and Science has developed certified reference materials (concrete, soil, and metal radioactive liquid) for measuring gamma-emitting radionuclides to improve and maintain the quality assurance and quality control of the radioactivity measurement in decommissioning nuclear power plants. The raw materials that make up each CRM were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on the radionuclides in the CRMs fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. In concrete analysis, the reference values for five radionuclides (60Co, 241Am, 134Cs, and 137Cs) in the CRM were in the range of 15-40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In soil analysis, the reference values for the 137Cs and 60Co were 118.7 and 124.4 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In metal radioactive liquid analysis, the reference values for 134Cs, 137Cs and 60Co in the CRM were in the range of 200-270 Bq/kg, and the expanded uncertainty was within 7% (k = 2).

The decommissioning of domestic Nuclear Power Plants (NPPs) in Korea is expected to begin with the Kori-1, which was permanently shutdown in 2017. In addition, Wolsong-1 has been also permanently shutdown, and another type will be the decommissioning project following Kori-1. KHNP is promoting operation and decommissioning projects as the owner of NPPs, and the Central Research Institute (CRI) has been developing a Final Decommissioning Plan (FDP) for the decommissioning license document. The FDP consists of 11 major chapters in the order of overview of the project, characteristic evaluation, safety assessment, radiation protection, decontamination & dismantlement activities, waste management, etc. The contents described in each chapter are individual chapters, but there are also parts that consider the connection with other chapters. The CRI, which develops the FDP for the first decommissioning project in Korea, has spent a lot of time and effort considering this and has been proceeding through trial and error until the present stage. Therefore, this study aims to explain the current status of FDP, a license document for domestic decommissioning projects, and the link between major input data in major chapters. It can be said that System, Structure, and Components (SSCs) subject to dismantling are considered as the scope of FDP. Chapters that perform estimations on these dismantling targets may include safety assessments, exposure dose assessments for workers and residents, and waste inventory assessments. Therefore, an important part of performing the estimation works is to consider the entire scope of decommissioning activities, and as a way, it can start from data based on the inventory data. After generating the inventory data, the waste treatment classification for the inventory is designated by reflecting the results of the characterization. In addition, for cost estimation, the cost of decommissioning project is predicted by inputting some data (i.e., UCF) such as work process, number of workers, and time required for each item with data reflected in quantity and characterization. After that, based on these inventory, characterization, and UCF data, accident scenarios and industrial safety evaluation are performed for the safety assessment. The worker exposure dose is estimated by considering the dose rate of the workspace with these data. In the case of the amount of waste, the final amount of waste is estimated by considering the factors of reduction and decontamination. In summary, the main estimation contents of FDP are evaluated by adding elements required for the purpose of each chapter from data combined with inventory, characterization, and UCF, so the contents of these chapters are based on the logic of considering the entire scope of decommissioning in common.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

Kori Unit 1 was permanently shut down in 2017 and is currently being prepared for decommissioning. Decommissioning waste generated during the decommissioning of a nuclear power plant has the characteristic of being generated in large quantities over a short period. Therefore, if proper management is not carried out, abnormal situations (i.e., unauthorized disposal, diversion, etc.) may occur. According to IAEA General Safety Report Part 6, radioactive waste shall be managed for all waste streams in decommissioning. This means ensuring that all waste streams are managed by the recorded inventory of all decommissioning waste and verifying that the recorded inventory is reasonable. The radioactive waste management has been managed in units such as mass and radioactivity. However, in the case of decommissioning waste, the amount is very large, so management by radioactivity is expected to have limitations. Therefore, in this study, a simple test was conducted to verify the decommissioning waste generated by a hypothetical scenario by mass. In this study, establish a scenario assuming various flows of decommissioning waste expected to be generated and calculate the expected inventory of decommissioning waste using Microsoft Excel. Specifically, using “Material Unaccounted For” (MUF), a material balance equation in IAEA Services Series 15, Nuclear Material Accounting Handbook, the error inventory was calculated as the difference between the physical inventory of decommissioning waste in the area and the ending inventory. We propose a simple test scenario to verify the flow of decommissioning waste by verifying that the error inventory reasonably matches the set allowable error. This study aims to verify the inventory of decommissioning waste using the material balance methodology used for nuclear material accounting. It is expected that the safety and reliability of the nuclear power plant decommissioning process can be secured by verifying that the total inventory of equipment before decommissioning and the inventory of remaining equipment and decommissioning waste after decommissioning are reasonably consistent.

When decommissioning a nuclear power plant, it is expected that clearance or radioactive waste (e.g., soil, concrete, metal, etc.) below the low-level will be generated in a short period on a large scale. Among the various types of waste, most of the contaminated soil is known to be classified as clearance or the (very) low-level radioactive waste. Accordingly, an accurate measurement and classification of contaminated soil in real-time during the decommissioning process can efficiently reduce the amount of soil waste and the possibility of contamination diffusion. However, in order to apply a system that measures and classifies contaminated soil in real-time according to the level of contamination to the decommissioning site, a demonstration is required to evaluate whether the system is applicable to the site. In this study, to establish requirements for determining the applicability of the system to the decommissioning site, preceding cases from countries with abundant decommissioning experience were investigated. For example, MACTEC of the U.S. demonstrated the developed system at the Saxton nuclear power plant in the U.S. and confirmed that the amount of soil that can be analyzed per hour in the system is affected by radionuclides, minimum detectable activity (MDA), and applicable volume. In the future, therefore, we will utilize the result of this study to develop the requirements of demonstrating the system for measurement and classification of contaminated soil in real-time.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

For safe and successful decommissioning, it is one of the most important procedures that establishing the goal and complying with regulations of which final status of decommissioned site and building. The dose criteria for cyclotron facilities should be established and applied to reuse the site and building, since building and component of a cyclotron facility have been activated by incident secondary neutrons from radioactive isotope processes (e.g. 18O(p,n)18F, etc.). Furthermore, appropriate approaches should be applied to demonstrate compliance with the dose criteria for reliability of reuse. It is of noted that U.S. NRC (Nuclear Regulatory Commission) has confirmed that the residual radioactivity which distinguishable from background radiation results in a TEDE (Total Effective Dose Equivalent) does not exceed 25 mrem (0.25 mSv) per year as radiological criteria for unrestricted use of not only nuclear power plants but also cyclotron facilities referred to 10 CFR Part 20.1402. In addition, U.S. NRC noted the two approaches (i.e. dose assessment methods and, DCGL and final status surveys) which can be applied for demonstrating compliance with the dose criteria of 10 CFR Part 20 and recommended DCGL and FSS approach based on advantages and disadvantages of the two approaches. In order to using DCGL and FSS approach, U.S. NRC suggested screening approach; using DandD Version 2 which assesses TEDE under ICRP 28 and site-specific approach; using all models or computational codes which approved by NRC staff. There are several foreign cases that release of cyclotron facilities after decommissioning (i.e. U.S. and Japan). U.S., for examples, there are two DCGL approach cases and one dose modeling case based on 25 mrem per year same as reactor facilities. The dose modeling case, however, which may not be really used in Korea because of its low applicability. On the other hand, Japan case did not establish any radiological criteria for site and building reuse such as DCGL and just confirm “no more contamination” which is all residual radioactivity is lower than MDC based on real survey. Japan case also may not be used in Korea since criteria of “no more contamination” is not clear and hard to apply for all sites. Considering regulations and criteria for site release and reuse in Korea, this study aims to suggest radiological criteria and the demonstration approach of compliance for decommissioning of cyclotron facilities based on Nuclear Safety Acts and NSSC notices.

The purpose of this report is to provide a summary of the Phase 1 Final Status Survey (FSS) Final Report results and overall conclusions which conduct that the Zion Nuclear Power Station (ZNPS) facility and site meets the 25 mrem(0.25 mSv)per year release criterion as established in Nuclear Regulatory Commission Regulation (NRC) 10 CFR 20.1402 “Radiological Criteria for Unrestricted Use”. The FSS results provided assessment and summarize that any residual radioactivity results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group (AMCG) that does not exceed 25 mrem per year, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). The release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary. ZionSolutions, a decommissioning service provider, estimates that a total of four (4) FSS Final Reports be generated and submitted to the NRC during the decommissioning project. ZionSolutions established the Characterization/License Termination (C/LT) Group, within the Radiation Protection division, with sufficient management and technical resources to fulfill project objectives. The C/LT Group is responsible for the safe completion of all surveys related to characterization and final site closure. Approved site procedures and detailed Technical Support Documents (TSD) direct the FSS process to ensure consistent implementation and adherence to applicable requirements. The development and planning phase was initiated in 1999 by the “Zion Station Historical Site Assessment” (HSA) and the initiation of the characterization process for FSS. Develop the information necessary to support FSS design, including the development of Data Quality Objectives (DQOs) and survey instrument performance standards. DQOs are qualitative and quantitative statements derived from the DQOs process that clarify technical and quality objectives. The next step, FSS design utilizes the combination of traditional scanning surveys, systematic sampling protocols and investigative/judgmental methodologies to evaluate survey units relative to the applicable release criteria for open land sample plans. To aid in the development of an initial suite of potential radionuclides of concern for the decommissioning of ZNPS, the analytical results of representative characterization samples collected at the site were reviewed. At this FSS design step, the Radionuclides of Concern (ROC) are determined. As Co-60 and Cs-137 account for 99.5% of the analysis results of concrete core sampling data form ZNPS’s Containment Building and Auxiliary Building, they are determined and used as the basic ROC in the survey design. Additionally, site information is described and Historical Site Assessment (HSA) is performed. Data collected for the initial HSA will be used to establish the initial regional survey unit and corresponding MARSSIM classification. Next, an assessment of the collected data is performed using the DQO process, and a survey methodology is established by selecting a sampling method and measuring instrumentation. These result judgments provide guidance for C/LT Engineer to interpret findings using the Data Quality Assessment (DQA) process, which analysis Recorded data, Missing values, Deviation from established procedure, and Analysis flags. In conclusion, FSS is the process used to demonstrate that the ZNPS facility and site comply the radiological criteria for unrestricted use specified in 10 CFR.20. The purpose of FSS Sample Plan is to describe the methods to be used in planning, designing, conducting, and evaluating the FSS.

Wolsong Unit 1 nuclear power plant, which was permanently shut down in 2019, has a 678 MWe calandria vessel of the CANDU-6 type pressurized heavy-water reactor model. The calandria inside the vault is a horizontal cylindrical vessel made of stainless steel with a length of 7.8 m and a thickness of 28.6 mm. For the entire dismantling processes of a nuclear power plant, dismantling works cannot be performed using only one cutting technology and method, and when performing dismantling of a calandria vessel, various systems and components can be used for cutting and dismantling. The calandria vessel is located in a concrete compartment called a vault, and in order to safely dismantle the calandria vessel, the spread of radioactive contaminants from inside of the vault to the outside must be prevented. We designed dismantling processes using the laser cutting method to dismantle the calandria vessel and end shields. We must minimize the risk of internal radiation exposure to workers from aerosols derived from the thermal cutting processes. Therefore, we need a way to prevent secondary contamination from spreading outside the vault and within the reactor building. The path through which radioactive contaminants move is that the flying airborne products generated during the cutting process inside the vault where the calandria is located do not stay in place but spread outward through the opening of the RM-Deck structure at the top. Therefore, facilities or devices are needed to effectively prevent the spread of radioactive contaminants by blocking the expected movement path. By using these facilities or devices, it is possible to prevent the movement of radioactive aerosol particles between the location of the worker and the location of the cutting area where the calandria is located, thereby preventing internal exposure through the worker’s breathing. In addition, by using these, the cutting area where airborne pollutants are generated can be designed as an isolated work space to prevent the spread of radioactive contaminants. In this study, we propose a method of facilities for confining radioactive aerosol particles and preventing the spread of contamination when thermal cutting of the calandria vessel within the vault.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of Metal radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF). Since a large quantity of metal radioactive waste is generated during the decommissioning of a nuclear power plant, the application of a metal melter for reduction is considered. The metal waste is heated to a temperature above the melting point and separated into liquid and gas forms. Nuclides existing on the surface of metal waste vaporize in a melting furnace to become dust or collect in sludge. Nonvolatile nuclides such as Co, Fe and Mn remain in ingot, but other nuclides can be captured and reduced with dust and sludge. And the types of melting furnaces to be applied can be broadly classified into Atmospheric Induction Melter (AIM) and Vacuum Induction Melter (VIM). Therefore, this review intends to compare the two types of metal furnaces to be included in RWTF.

Large amounts of concrete, metal, soil, and other radioactive waste are generated not only from nuclear power plants operating in Korea but also from nuclear power plant decommissioning. If it is confirmed through measurement of residual radioactivity that the concentration is below the allowable clearance level, they can be managed as general or industrial waste in accordance with the Nuclear Safety Act. The Korea Radioactive Waste Agency predicts that very low-level radioactive waste will be generated the most, at about 67.1%. If waste below clearance level among very low-level radioactive waste can be evaluated and reduced, a lot of costs can be saved. Among radioactive wastes, metal wastes in particular have various sizes, shapes, and densities. If radioactivity is measured without properly considering this, a large error occurs in the measured value even if the radioactivity value is the same. This requires a conservative measurement method using density correction taking into account the self-absorption effect. For conservative measurements, it is essential to compare measured values with calculated values using MCNP6 (Monte Carlo N-Particle). You must enter the geometry of the measurement environment and derive calculated values using F8 Tally. Clearance level of radioactive waste is determined through the above method. In addition, sufficient MDA (Minimum Detectable Activity) must be secured to determine clearance level by using NaI(Tl), plastic scintillator configuration, and lead shielding. Nuclide analysis is performed using a NaI(Tl) scintillator and the total gamma radioactivity is evaluated using a highly efficient plastic scintillator.

The Radiation and Decommissioning Laboratory of Central Research Institute (CRI) of Korea Hydro and Nuclear Power Co. (KHNP) performs research to technically support the effective management of radiological hazards to avoid risks to civilians, the workers, and the environment from the radiological risks. The laboratory mainly consists of three technical groups: decommissioning and SF technology group, radiation and chemistry group, and radwaste and environment group. The groups carry out various R&D such as decommissioning, spent fuel management, radiation protection, water chemistry management, and radioactive waste management. The laboratory also technically supports the calibration of radiometric instruments as a Korea Laboratory Accreditation Scheme (KOLAS), approval for decommissioning, guidance for radioactive waste management, state-of-the-art technology evaluations, and technology transfer.

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.