The nuclear power plant (NPP) decommissioning market is expected to expand not only domestically but also overseas. Proven technologies must be applied to decommission NPP. This is based on Article 41-2, Paragraph 2 of the domestic ‘Enforcement Decree Of The Nuclear Safety Act’. Proven technology refers to technology that has verified that it can be applied in the field through demonstration. In other words, in order to carry out NPP decommissioning, verification must be done. Demonstration refers to reducing technological uncertainty and directly verifying services implemented in the field. From a technology commercialization perspective, demonstration requires an approach based on technology readiness level (TRL) from a technology perspective and market readiness level (MRL) from a market perspective. The characteristics of demonstration also differ depending on the characteristics of each field. The demonstration in the field of nuclear energy is the demonstration of demand matching. This is to confirm the feasibility of the technology in the company’s required environment. In order to perform demonstration, a scenario must be derived by reflecting demonstration design considerations. After evaluating the derived scenario, an actual assessment is conducted using lab-based demonstration/virtual environment demonstration/real environment demonstration. What must be preceded by an actual assessment is confirming the consumer’s requirements. In this study, the necessary environment and requirements of consumer’s to perform NPP decommissioning were reviewed. The domestic decommissioning procedure requirements management system presents decommissioning procedures, potential worker accidents, and worker requirements. In the case of foreign countries, it was confirmed that complex wide need, cost benefit, risk reduction, waste generation, operation, reliability and maintenance (RAM) improvement and quantitative measures were evaluated for the technology to be demonstrated. Also the requirements for demonstrating decommissioning need to a detailed review of actual decommissioning cases. Therefore, a comparison must be made between the requirements based on actual NPP decommissioning cases and the requirements derived from this research process. Afterwards, the empirical research approach proposed by the Ministry of Trade, Industry and Energy was applied. The empirical research approach proposed by the Ministry of Trade, Industry and Energy is to secure a track record over a certain period of time and performance under conditions similar to the actual environment in the final research stage at the TRL level 6 to 8. Through this, it will be possible to confirm the suitability of overseas technology for domestic application.

Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes generated from nuclear power plant decommissioning. It is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. The contamination evaluation module of the system has two sub modules. One is for quick measurement with NaI (Tl) detector and the other is for accurate measurement with HPGe detector. The container used at the system for wastes handling has capacity of 100 kg and made of stainless steel. According to the measurement result of Co-60 and Cs-137, the waste is classified as waste for disposal or waste for clearance. Performance of the system was demonstrated using RM (Reference Material) radiation source. This year, necessity of system improvement was suggested due to revised operation requirements. So, the system should show throughput of more than 1 ton/hr and Minimum Detectable Activity (MDA) of less than 0.01 Bq/g (1/10 of criteria for regulatory clearance) for Co-60 and Cs-137. And soil waste become main target of the system. For this, the container used for soil waste handling should have capacity of 200 kg. As a result, material for the container need to be changed from stainless steel to plastic or FRP (Fiber Reinforced Plastics). And large area detector should be introduced to the system to enhance processing speed of the system. Additionally, container storage rack and conveyor system should be modified to handle 200 kg capacity container. Finally, moving path of the container will be redesigned for enhanced throughput of the system. In this paper, concept development of the system was suggested and based on that, system development will be followed.

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 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.

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.

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.

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

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

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

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 liquid radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF) or on-site. In this regard, there is no Code and Standard for the amount of liquid radioactive waste generated during NPP are dismantled, but ANSI/NS-55.6 describes the amount of liquid radioactive waste generated from a light water reactor type NPP. This code is applied to nuclear power-related facilities such as domestic NPP and radioactive waste disposal facility. Therefore, this review intends to suggest an application plan for domestic NPP decommissioning through codes for liquid radioactive waste expected to generate during nuclear power plant decommissioning.

On March 11 2011, Fukushima Daiichi nuclear power plant site was attacked by a huge tsunami caused by Tohoku Pacific Ocean earthquake. Nuclear fuels of unit 1, 2, and 3 of Fukushima Daiichi nuclear power plant was melted down by the disaster. After the accident, Japan’s government has announced “Mid-and-Long-Term Roadmap towards the decommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Station Units 1-4”. The topics of roadmap is made of measures to deal with contaminated water, removal of fuel rod assemblies from spent fuel pools, retrieval of fuel debris, measures to deal with waste materials, and other operations. To support the activity of the roadmap, various facilities about decommissioning have been established and operated on inside or outside of Fukushima Daiichi nuclear power plant site. Representatively, Collaborative Laboratories for Advanced Decommissioning Science which conducts R&D decommissioning, Naraha Remote Technology Development Center which develops remotes robots and VR (Virtual reality), Okuma Analysis and Research Center which performs radiochemical analyses for radioactive waste, and Fukushima Environmental Safety Center which conducts environmental dynamics and radiation monitoring.

Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

Under Article 17 of the Radioactive Waste Management Act and Article 12 of the Enforcement Decree of the Radioactive Waste Management Act, KHNP shall reserve the cost for the decommissioning of NPPs as provisions. To preserve the value, an additional amount considering the discount rate is to be added annually. The initial provision is decided by estimating the decommissioning cost of NPP at the time of commercial operation, calculating the future cost by applying the inflation rate to the expected start date of decommissioning, and then discounting it at a discount rate to the present value. According to the current notice, the period for applying inflation and discount rate is defined as the period of 5 years added to the design life of NPP, which is presumed to be due to the assumption that all decommissioning costs are incurred at once 5 years after the permanent shutdown of the power plant. However, assuming that the actual decommissioning period of a domestic nuclear power plant is generally planned for 15 years, it can be expected that most of the decommissioning activities will begin after the decommissioning preparation and transition period, or 5 years after permanent shutdown of the plant. Considering this, it can be said that the current period (5 years + design life) for applying inflation and discount rate is set a little conservatively. In this paper, the initial provision is calculated by appropriately distributing the decommissioning costs of overseas NPPs categorized by International Structure for Decommissioning Costing (ISDC) during the planned decommissioning period of domestic NPPs, and then adding up the decommissioning cost each year by separately applying the inflation and discount period, which was compared with the results calculated using the current method. Through this, it was confirmed that the revised method had the effect of reducing the initial provision by 2.2% to 5.7% compared to the current method depending on the gap between inflation rate and discount rate, which can be converted to about 8 years of inflation and discount period used in the current method. It is expected that this paper will be used in the future as a basic reference for developing a more accurate method for calculating the initial provision of decommissioning cost.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

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.

Minimizing of radiation exposure for the operating and decommissioning personnel is a key indicator for safe operation of the NPP. This is reflected in the application of the ALARA (As Low As Reasonable Achievable) principle. The main objectives of radiation management during full system decontamination for NPP decommissioning are to reduce the exposure dose, prevent contamination of the body and reduce solid radioactive waste. In order to reduce exposure of workers, the dose rate should be reduced by installing a temporary shield after evaluating the dose rate for the piping, component and decontamination equipment of the decontamination path before full system decontamination. Furthermore, unnecessary exposure to radiation should be reduced by thoroughly entering and exciting the radiation area and limiting the access to the high-radiation area except for workers or persons concerned. A telemetric dosimetry system should be as installed to remotely monitor radiation levels at different locations within the decontamination flow path. Remote monitoring of radiation fields using teledosimetry worked well in assessing process effectiveness and is highly recommended. However, care must be taken to place the detectors in appropriate locations. For the prevent of body contamination, it is necessary to install a fence using a heat-resistant waterproof sheet to prevent leakage of highly radioactive contamination water. When replacing high-dose filters and ion exchange resins, it is necessary to remotely monitor to reduce the exposure dose of workers.

Waste containers for packaging, transportation and disposal of NPP (Nuclear Power Plant) decommissioning wastes are being developed. In this study, drop tests were conducted to prove the safety of containers for packaging of the wastes and to verify the reliability of the analysis results by comparing the test and analysis results. The drop height of the waste containers was considered to be 30 mm, which is the maximum lifting speed of a 50 tons crane in the waste treatment facility converted to the drop height. Drop orientation of the containers was considered for bottom-end on drop. The impact acceleration and strain data were obtained to verify the reliability of the analysis results. Before and after the drop tests, measurement of the dose rate and the radiographic testing for concrete wall, and measurement of the wall thickness of steel plate were conducted to evaluate the radiation shielding integrity. Also, measurement of bolt torque, and visual inspection were conducted to evaluate the loss or dispersion of radioactive contents. After the drop tests, the radiation dose rate on the container surface did not increase by more than 20%, and there was no crack in the concrete. In addition, the thickness of the steel plate did not change within the measurement error. Therefore, the radiation shielding integrity of the container was maintained. After the drop tests, the lid bolts were not damaged and there was no loss of pretension in the lid bolts. In addition, there was no loss or dispersion of the contents as a result of visual inspection. In order to prove the reliability of the drop analysis results, safety verifications were performed using the drop test results, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, the structural integrity of the waste containers was maintained under the drop test conditions.