The decommissioning of Kori Unit 1 is expected to generate a large amount of clearance waste. Disposing of a large amount of clearance waste is economically costly, so a recycling method has emerged. However, clearance waste recycling is expected to cause many conflicts among various stakeholders. In the previous study, possible conflicts were selected in consideration of the domestic environment and major issues. Based on this, this study classifies stakeholders involved in conflicts by group, and suggests ways to enhance understanding by stakeholder and enhance reliability. In this study, stakeholders are classified into four groups that share the same conflicts, and each of the following measures is suggested. 1) Stakeholder Engagement. 2) Common understanding of radiation risks, dialogue between the public/recycling industry/ regulatory agency. 3) Incentives to promote recycling clearance waste. 4) Reliable outlet store for recyclable clearance waste. The above understanding enhancement measures are presented so that a solution to conflict can be smoothly derived when designing a clearance waste-related consultative body composed of interested parties in the future. As a more specific solution, measures to enhance stakeholder trust can be suggested for each understanding enhancement measure. Reliability enhancement measures are also presented so that they can be applied to each stakeholder group, and these are as follows. 1) Write a stakeholder engagement plan, Measures for stakeholder participation in measuring the radioactivity concentration of clearance waste. 2) Active use of easy-to-understand radioactivity comparison data, Expansion of information on environmental radiation dose to public, nuclear/radiation education, Held a tour event at the nuclear power plant decommissioning site, New website for clearance waste information disclosure. 3) Incentives for recycling industries in which the Ministry of Environment or KHNP partially bears the losses that occur when the sales rate is low. Incentives are provided to consumers by including recyclables of clearance waste for Green Card’s green consumption points. 4) Online outlets open for recyclable clearance waste with easy-to-understand radioactivity comparison data. It is expected that if the above-mentioned reliability enhancement measures are used, it will be possible to secure the trust of stakeholders and reduce the gap between stakeholders in the future clearance-related consultative body.

Kori unit 1 was permanently shut down in 2007 and is currently awaiting approval for decommissioning and dismantling (D&D). The wastes generated during decommissioning is estimated to be approximately 14,500 of 200 L drums. In this study, the treatment process of decommissioning wastes will be reviewed through the case of the US Zion nuclear power station (ZNPS). Zion unit 1 and 2 received an operating license in 1973 and were permanently shut down and the spent nuclear fuel was transferred to the pool in 1998. The decommissioning was carried out according to the following five steps; (1) safe storage (SAFSTOR) dormancy, (2) preparation for decommissioning, (3) establishment of independent spent fuel storage installation (ISFSI) and transfer of the spent fuel and greater than class C radioactive materials, (4) decommissioning operations and (5) site restoration. The total volume of waste generated during decommissioning was expected to be approximately 1.7×105 m3. This is far above the Kori unit 1 waste estimation because ZNPS has a history of accidents and includes soil waste. Wastes were treated differently according to their properties and locations.

Reliable evaluation of radioactivity inventory for the nuclear power plant components and residual materials is very important for decontamination and decommissioning. This can make it possible to define optimum dismantling approaches, to determine radioactive waste management strategies, and to estimate the project costs reasonably. To calculate radioactivity of the nuclear power plant structure, various information such as interest nuclide, cross-section, decay constant, irradiation time, neutron flux, and so on is required. Especially irradiation time and neutron flux level are very changeable due to cycle specific fuel loading pattern, the plant overhaul, cycle length. However most of the radioactivity calculations have generally been performed assuming one representative or average neutron flux during the lifetime of the nuclear power plant. This assumption may include excessive conservatism because the radioactivity level has the characteristics of saturation and decay. Therefore, considering these variables as realistically as possible could prevent overestimation. In order to perform realistic radioactivity calculation, we developed monthly relative power contribution factor applying plant-specific operation history and cycle-specific neutron flux. The factors were applied to the radioactivity calculation. The calculation results ware compared with measured values of the neutron monitors that were actually installed and withdrawn from the nuclear power plant. As a result of the comparisons, there are good agreements between the calculated values and measured values. These accurate calculation results of radioactivity could contribute to the establishment of radioactive waste dismantling strategies, the classification of radioactive waste, and the deposit of disposal costs for safe and reasonable decommissioning of the nuclear power plant.

Radiological characterization is important in decommissioning and dismantling of nuclear facilities, in order to assess the radioactivity concentration, classify the wastes, and secure workers’ safety. The Some components such as Reactor Pressure Vessel (RPV) in nuclear facilities has dose rate higher than Sv/hr, thus in-situ gamma spectroscopy systems suffer from a very high count rate which causes energy resolution degradation, photo-peak shift, and count loss by pile-up and dead-time. The system must be operated in a very high count rate, in order to measure spectra precisely and to quantify radionuclide contents. In order to apply in-situ measurement in high radiation dose rate environment, the sensor, front-end electronics, and data acquisition (DAQ) should be carefully selected and designed as well as precise design of collimators and radiation shield. In this paper, the components of the detector system were selected and performance was evaluated in a high count rate before design the collimator and shield. A LaBr3 coupled with a PMT having short decay time constant (16 nsec) was selected for high count rate application, and two different amplifiers (a conventional charge sensitive preamplifier with 50 usec decay time constant, and wide-band voltage amplifier) were tested. As DAQs, DT5781 (14 bit, 100 MS/s, CAEN) of Pulse Height Analysis (PHA) which is conventionally used signal processing method in the gamma spectroscopy, and DT5730 (14 bit, 500MS/s, CAEN) of Pulse Shape Discrimination (PSD) which is similar to Charge to Digital Convertor (QDC) were used. The number of photons incident to the detector was varied by changing the detector-source distance with Certificate Radiation Material (CRM), and compared to the output count rate. The count rate capability, and energy resolution with different amplifier and DAQ was evaluated. Additionally, the performance of DAQs in extremely high count rate was evaluated with signal data generated by the emulator which can simulate the detector signal waveforms fed into the DAQ based on the measured spectrum.

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

The safe, efficient and cost-effective decommissioning and dismantling of radioactive facilities requires the accurate characterization of the radionuclide activities and dose rate environment. And it is critical across many nuclear industries to identify and locate sources of radiation accurately and quickly. One of the more challenging aspects of dealing with radiation is that you cannot see it directly, which can result in potential exposure when working in those environments. Generally, semiconductor detectors have better energy resolution than scintillation detectors, but the maximum achievable count rates are limited by long pulse signals. Whereas some high pure germanium detectors have been developed to operate at high count rates, and these HPGe detectors could obtain gamma-ray spectra at high count rates exceeding 1 Mcps. However, HPGe detectors require cooling devices to reduce the leak currents, which becomes disadvantageous when developing portable radiation detectors. Furthermore, chemicalcompound semiconductor detectors made of cadmium telluride and cadmium zinc telluride are popular, because they have good energy resolution and are available at room temperature. However, CdTe and CZT detectors develop irradiation-induced defects under intense gamma-ray fields. In this Review, we start with the fundamentals of gamma rays detection and review the recent developments in scintillators gamma-ray detectors. The key factors affecting the detector performance are summarized. We also give an outlook on the field, with emphasis on the challenges to be overcome.

A large spectrum of possible stakeholders and important factors for safety improvement during decommissioning of nuclear facilities should be identified. Decommissioning includes additional aspects which are of interest to a wider range of stakeholders. The way in which local communities, the public in general, and a wide range of other parties are engaged in dialogue about decommissioning of nuclear facilities is likely to become an increasingly important issue as the scale of the activity grows. Timely stakeholder involvement may enhance safety and can encourage public confidence. Stakeholder engagement may result in attention to issues that otherwise might escape scrutiny. Public confidence is improved if issues that are raised by the public are taken seriously and are carefully and openly evaluated. Experience in many countries has shown that transparency can be an extremely effective tool to enhance safety performance. It sets out the development and implementation of an effective two-way process between the organization and stakeholders. Meaningful engagement is characterized through a flow of communication, opinions and proposals in both directions and the use of collaborative approaches to influence and explain decisions. The process is one in which an organization learns and improves its ability to perform meaningful stakeholder engagement while developing relationships of mutual respect, in place of one-off consultations. The evolving nature of this process is particularly relevant to pipeline projects, which will have differing stakeholder engagement requirements at each phase of the project lifecycle. Activity undertaken at all stages of the process should be documented to ensure engagement success can be reviewed and improved and to ensure historical decisions or engagements are captured in case stakeholders change during the progression of time and previous consultation records are required.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

Laser cutting has been attracting attention as a next-generation tool in application for nuclear decommissioning. It enables high-speed cutting of thick metal objects, and its narrow kerf width greatly reduces the amount of secondary waste compared to other cutting methods. In addition, it only requires the relatively small cutting head without any complicated equipment, and long-distance cutting apart from a laser generator is possible using beam delivery through optical fiber. And there is almost no reaction force because it is non-contact thermal cutting. For these reasons, the laser cutting is very advantageous for remote cutting. In laser cutting, the irradiated laser power is absorbed and consumed to melt the material of the cutting target. When the applied laser power is greater than the power consumed for melting, the residual power is transmitted to the back of the cut object. This residual power may unintentionally cut or damage undesired objects located behind the cutting target. In order to prevent this, it is necessary to adjust the laser power for each thickness of the target object to be cut, or to increase the distance between the cut target and the surrounding structures so that the transmitted power density can be sufficiently lowered. In this work, safety study on residual power that penetrates laser-cut objects was conducted. Experimental studies were performed to find safe conditions for irradiation power density that does not cause surface damage to the stainless steel by adjusting the laser power and stand-off distance from the target.

Kori unit 1, Korea’s first light-water nuclear power plant, was permanently shut down in June 2017. The operator, Korea Hydro & Nuclear Power Co. (KHNP), submitted a final dismantling plan for Kori unit 1 to the Nuclear Safety and Security Commission (NSSC) in May 2021. Pursuant to this procedure, the NSSC is preparing regulations for the decommissioning stage of large nuclear facilities for the first time in the Republic of Korea. The Korea Institute of Nuclear Non-proliferation and Control (KINAC) is also considering applying regulations on safeguards. Moreover, the International Atomic Energy Agency (IAEA) developed the “International Safeguards Guidelines for Nuclear Facilities under Decommissioning” in 2021. The guidelines describe the detailed application of safeguards measures to be considered when decommissioning nuclear facilities, dismantling essential equipment, and providing relevant information to the IAEA, as well as the scope of IAEA inspections. In addition, Dr. R. Bari of the Brookhaven National Laboratory (BNL) proposed the Facility Safeguardability Assessment (FSA), a methodology that reflects facility characteristics from the design stage to ensure that designers, national regulators, and the IAEA communicate smoothly regarding safeguards measures. The FSA process derives expected problems with safeguards measures considering new nuclear facilities by analyzing the gap of safeguards measures applied to existing similar nuclear facilities. This study uses the existing FSA methodology to predict problems related to safeguards measures when decommissioning nuclear facilities and to analyze deviations from safeguards measure requirements according to IAEA guidelines. To this end, the reference facility is set as an operating pressurized light water reactor; the issues with the safeguards measures are summarized using the FSA Process; and a draft safeguards concept for nuclear facilities under decommissioning is designed. Furthermore, validity is confirmed through a simple analysis of the diversion path, and implications and lessons are derived. Through this, it is possible to anticipate new safeguards measures to be applied when decommissioning nuclear facilities in the Republic of Korea and review problems and considerations in advance.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’ radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

Considering the characteristics of nuclear power plants in order to decommission nuclear power plants safely and economically, this thesis provides a methodology for optimizing the technology for developing decommissioning characteristic evaluation system using simulation technology for core facilities of the plants based on 3D that reflects various factors. The results of pollution assessment and radiation assessment for the Kori Unit 1 reactor building, auxiliary building, and each major device are displayed in 3D drawings and viewer, and the radiation dose rate and radiation assessment results are displayed separately for each major location. Furthermore, this D/B development method which includes inserting result values of characteristic evaluation and the quantity of waste is one of the main technology to optimize the system which enables users to select decommissioning processes and predict the quantity of waste. (Refer to the presented 3D models of the containment building, D/B, tag search module, the scale calculation result of models after visualizing the result value of 3D based decommissioning characteristic evaluation) The methodology for optimizing decommissioning characteristic evaluation result value DB development system using 3D models of the first major nuclear power plant allows the display of decommissioning characteristic values in virtual reality, the selection of decommissioning process, the establishment of the decommissioning procedure. Hence, this study is expected to provide reliable guidelines for managing a decommission business efficiently in the near future and can be used in the related field if needed.

The Kori Unit 1 and Wolsong Units 1, commercial reactors in South Korea, were permanently shut down due to the expiration of their design lifetime. Therefore, nuclear power plants that have been permanently shut down must be dismantled, and the site must be finally released after removing the remaining radionuclides. Domestic regulatory standards for site remediation should not exceed 0.1 mSv per year based on effective dose. In addition, it is necessary to calculate the preliminary Derived Concentration Guideline Levels (DCGL) to prove that the conditions are met. Therefore, in this study, the input factor considering the geological characteristics of the site of Kori Unit 1 was investigated, and the preliminary Derived Concentration Guideline Levels were calculated and compared with the results of previous studies. As a result of comparative analysis, 60Co, 134Cs, and 137Cs, which are gamma-ray emitting radionuclides, had similar values to DCGL of previous studies A and B. However, 63Ni, a beta-rayemitting nuclide, was 5.94×104 Bq·g−1 in this study and 8.47×101 Bq·g−1in previous study B, resulting in a difference of about 700 times. In addition, in the case of 90Sr, this study and previous study A were derived similarly, but this study was 5.34×101 Bq·g−1 and previous study B was 1.18×10−1 Bq·g−1, resulting in a difference of about 450 times. This difference is judged to be because, unlike this study using only the industrial worker scenario, in the case of previous study B, the resident farmer scenario was mixed and used, which considers the internal exposure caused by ingestion of food produced in the contaminated area. In this study, it was confirmed that DCGL according to the change of geological factors of the site did not have a significant effect on gamma-ray-emitting nuclides. However, it was confirmed that considering the intake of food affects the DCGL of beta-ray-emitting nuclides. Therefore, there is a need to conduct future studies applying intake input factors that meet domestic conditions.

Radiological characterization, one of the key factors for any successful decommissioning project for a nuclear facility, is defined as a systematic identification of the types, quantities, forms, and locations of radioactive contamination within a facility. This characterization is an essential early step in the development of a decommissioning plan, in particular during transition period after permanent shutdown of the facility, and also to be used for classification of decommissioned radioactive wastes so that their disposal criteria can be met. Therefore, the characterization should be well planned and performed. In the transition period, the characterization information developed during the operational phase is usually reexamined with respect to the applied assumptions, the actual status of the facility after shutdown, the accuracy of the required measurements and changes in its radiological properties to support the development of the final decommissioning plan. Based on some national (Korean, USA’s and Japanese) laws including the related regulations, and some related documents published by OECD/NEA, IAEA, and ASTM, key elements of radiological characterization, which should be developed in the transition period, could be proposed as the followings. The key elements might be an operational history including facility operation history and contamination by events and/or accidents, radiological inventory of the facility and site area, characterization survey including in-situ survey and/or sampling and analyses, radiological mapping (which is able to identify radiological contamination levels of SSCs, and the facility area and, if contamination may be suspected, the surroundings) with tabulating, residual radioactivity (or derived concentration guideline levels) of selected major radionuclides for remediation of the site, (retainable and retrievable) recording, and quality control and quality assurance. In review process of the operational history, interviews of current or former long-tenured knowledgeable employees of the facility should be conducted to identify conditions that may have been missing from the records.

The decommissioning of nuclear power plant (NPP) generates large amount of waste. Since the most of the concretes are slightly surface contaminated, the accurate characterization and regionspecific surface decontamination are important for the efficient waste management. After the effective surface decontamination and separation, most of the concrete waste from decommissioning of NPP can be classified as a clearance waste. Various surface characterization and decontamination technologies are suggested. The mechanical technologies are simple and offers direct application. The laser-based technologies offer efficient separation and surface contamination. The high price, however, hesitates the application of the process. The nitro-jet technology, which is based on the evaporation of liquid nitrogen, allows the effective decontamination. However, the high price and uncertainty of large are application hinders the practical application in NPP decommissioning. In this paper, various technologies for characterization, handling, treatment, etc., will be discussed. The advantages and disadvantages of the technologies will be discussed, in terms of practical applications.

The establishment of processes for the decommissioning a Nuclear Power Plant (NPP) is one of the objects that must be prepared in carrying out the decommissioning project. In particular, in the domestic situation, where there is no experience of decommissioning commercial NPPs, it is necessary to organize the tasks and contents well in advance for the successful initiation of the project. Therefore, this study intends to present a guide-level approach to develop management for domestic decommissioning projects. As a documented template for recognizing a process, there may be a process map and description, and information such as the work structure and the relations between the activities should be indicated. In reality, activities will be managed through a set of computer system, so it would be better if the work content, activity flow, relation, management target information, computerization contents, etc. were materialized in the process. What is important here is to define the management areas and activities and draw the activity flow. Domestically, it has rich experience in construction of NPPs and has a track record of exporting NPPs to the UAE. From these experiences, we have established a framework for standardized work in construction management and construction processes, and are performing them through a computerized system. Since the work of decommissioning has a similar nature to that of construction, we will be able to benchmark the procedure for the decommissioning from the construction management procedures. Typically, in the case of schedule management, the concept and structure of the construction process will be applicable to the decommissioning. Meanwhile, the licensee of domestic decommissioning is the same as the licensee that performs the operation, and the members who will perform the decommissioning also have experience working in the operation period. Therefore, the decommissioning works are an extension of the task during operation. Representatively, there are some processes that can be applied as it is even when decommissioning, such as dismantling work and the safety management process of the radiation zone. Therefore, in carrying out the decommissioning of NPPs in Korea, processes and activities of the management area should be established from the construction processes with abundant experience and the processes during operation. Rather than making a completely new work process, this approach that properly reflects the existing work flow is expected to be an appropriate way to avoid the repulsion of employees and maladjustment to the new environment.

As Kori-1 permanently shut down in Korea, it is expected that a large amount of radioactive waste will be generated during decommissioning of nuclear power plants. Radioactive concrete waste is contaminated up to depth of 100 mm with radionuclides such as 137Cs and 60Co. The radioactive waste should be accurately classified to reduce the cost of disposing of radioactive waste. Therefore, the specific radioactivity of waste must be precisely evaluated by gamma-ray measurements emitted from the radionuclides. In general, the effectiveness of the radioactivity measurement and process is confirmed using certified reference material (CRM) composed of water or agar. However, the decommissioning waste differs from this CRM in apparent density and chemical elements, so the specific radioactivity is underestimated or overestimated. Therefore, reference material composed of the same apparent density and chemical elements as the sample is required to improve the quality of radioactivity measurement. The purpose of this study is to develop a concrete reference material for the nuclear decommissioning waste. The concrete reference material composed of SiO2, CaO, and Al2O3 were manufactured in compliance with ISO Guide 35. 10 bottles were randomly selected for homogeneity test, and 2 samples for analysis were taken from each bottle. The specific radioactivity was measured using an HPGe detector with an efficiency of 30%. The results of the homogeneity test of 137Cs and 60Co satisfied the requirements of ISO Guide 35. Coincidence summing and selfabsorption effects were corrected using the Monte Carlo efficiency transfer code and Monte Carlo NParticle transport code. The reference values of 137Cs and 60Co in the concrete reference material were evaluated in the range of 1,000–1,100 Bq·kg−1 and extended uncertainty was around 7%.

The remote dismantling system proposed in this paper is a system that performs the actual dismantling process using the process and program predefined in the digital manufacturing system. The key to the successful applying this remote dismantling system is how to overcome the problem of the difference between the digital mockup and the actual dismantling site. In the case of nuclear facility decommissioning, compensation between the virtual world and the real world is difficult due to harsh environments such as unsophisticated dismantling sites, radiation, and underwater, while offline programming can be proposed as a solution for other industries due to its sophisticated and controllable environment. In this paper, the problem caused by the difference in the digital mockup is overcome through three steps of acquisition of 3D point cloud in radiation and underwater environment, refraction correction, and 3D registration. The 3D point cloud is acquired with a 3D scanner originally developed in our laboratory to achieve 1 kGy of radiation resistance and water resistance. Refraction correction processes the 3D point cloud acquired underwater so that the processed 3D point cloud represents the actual position of the scanned object. 3D registration creates a transformation matrix that can transform a digital mockup of the virtual world into the actual location of a scanned object at the dismantling site. The proposed remote dismantling system is verified through various cutting experiments. In the experiments, the cutting test object has a shape similar to the reactor upper internals and is made of the same material as the reactor upper internals. The 105 successful experiments demonstrate that the proposed remote dismantling system successfully solved the key problem presented in this paper.