Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

Our research team has developed a gamma ray detector which can be distributed over large area through air transport. Multiple detectors (9 devices per 1 set) are distributed to measure environmental radiation, and information such as the activity and location of the radiation source can be inferred using the measured data. Generally, radiation is usually measured by pointing the detector towards the radioactive sources for efficient measurement. However, the detector developed in this study is placed on the ground by dropping from the drone. Thus, it does not always face toward the radiation source. Also, since it is a remote measurement system, the user cannot know the angle information between the source and detector. Without the angle information, it is impossible to correct the measured value. The most problematic feature is when the backside of the detector (opposite of the scintillator) faces the radiation source. It was confirmed that the measurement value decreased by approximately 50% when the backside of the detector was facing towards the radiation source. To calibrate the measured value, we need the information that can indicate which part of the detector (front, side, back) faces the source. Therefore, in this study, we installed a small gamma sensor on the backside of the detector to find the direction of the detector. Since this sensor has different measurement specifications from the main sensor in terms of the area, type, efficiency and measurement method, the measured values between the two sensors are different. Therefore, we only extract approximate direction using the variation in the measured value ratio of the two sensors. In this study, to verify the applicability of the detector structure and measurement method, the ratio of measured values that change according to the direction of the source was investigated through MCNP simulation. The radioactive source was Cs-137, and the simulation was performed while moving in a semicircular shape with 15 degree steps from 0 degree to 180 degrees at a distance of 20 cm from the center point of the main sensor. Since the MCNP result indicates the probability of generating a pulse for one photon, this value was calculated based on 88.6 μCi to obtain an actual count. Through the ratio of the count values of the two sensors, it was determined whether the radioactive source was located in the front, side, or back of the probe.

The increasing use of drones in terrorist attacks highlights the need for effective strategies to prevent and respond to drone terrorism. This study uses machine learning approach to identify factors that predict the success of drone terrorism and suggests policy alternatives for preventing such acts. Drone terrorism is becoming increasingly accessible due to advancements in information and communication technology, and events such as North Korea’s drone infiltration and the Russia-Ukraine war demonstrate the potential threat of drone attacks on Important National Facilities, including nuclear power plants. Using the Global Terrorism Database (GTD), this study analyzed drone terrorism incidents that occurred worldwide from 2016 to 2020. The study employed the Random Forest algorithm, which can incorporate multiple factors and their interactions, making it particularly suitable for social science research. The study provides new insights by deriving predictors that were previously overlooked in empirical analyses of drone terrorism. The findings of this study can aid in the establishment of anti-terrorism policies aimed at addressing the growing threat of drone terrorism. This can include the organization and expansion of the crisis management governance terrorism response council, the creation of a working manual through the partial revision of laws concerning drone terrorism response, and the implementation of anti-drone equipment and systems. Ultimately, the insights gained from this study can provide development of effective strategies aimed at preventing and responding to drone attacks. The study highlights the importance of proactive measures to mitigate the risks posed by drone technology in the context of terrorism.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

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.

In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

Disposal of radioactive waste requires radiological characterization. Carbon-14 (C-14) is a volatile radionuclide with a long half-life, and it is one of the important radionuclides in a radioactive waste management. For the accurate liquid scintillation counter (LSC) analysis of a pure beta-emitting C-14, it should be separated from other beta emitters after extracted from the radioactive wastes since the LSC spectrum signals from C-14 overlaps with those from other beta-emitting nuclides in the extracted solutions. There have been three representative separation methods for the analysis of volatile C-14 such as acid digestion, wet oxidation, and pyrolysis. Each method has its own pros and cons. For example, the acid digestion method is easily accessible, but it involves the use of strong acids and generates large amount of secondary wastes. Moreover, it requires additional time-consuming purification steps and the skillful operators. In this study, more efficient and environment-friendly C-14 analysis method was suggested by adopting the photochemical reactions via in-situ decomposition using UV light source. As an initial step for the demonstration of the feasibility of the proposed method, instead of using radioactive C-14 standards, non-radioactive inorganic and organic standards were investigated to evaluate the recovery of carbon as a preliminary study. These standards were oxidized with chemical oxidants such as H2O2 or K2S2O8 under UV irradiations, and the generated CO2 was collected in Carbo-Sorb E solution. Recovery yield of carbon was measured based on the gravimetric method. As an advanced oxidation process, our photocatalytic oxidation will be promising as a time-saving method with less secondary wastes for the quantitative C-14 analysis in low-level radioactive wastes.

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.

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

The nuclear power plant decommissioning project inevitably considers time, cost, safety, document, etc. as major management areas according to the PMBOK technique. Among them, document management, like all projects, will be an area that must be systematically managed for the purpose of information delivery and record maintenance. In Korea, where there is no experience in the decommissioning project yet, data management is systematically managed and maintained during construction and operation. However, if the decommissioning project is to be launched soon, it is necessary to prepare in consideration of the system in operation, what difference will occur from it in terms of data management, and how it should be managed. As a document that can occur in the decommissioning project, this study was considered from the perspective of the licensee. Therefore, the types of documents that can be considered at Level 1 can be divided into (1) corresponding documents, (2) project documents, (3) internal documents, and (4) reference materials. Four document types are recommended based on Level 1 for the classification of documents to be managed in the decommissioning of nuclear facilities. In this study, documents to be managed in the decommissioning project of nuclear facilities were reviewed and the type was to be derived. Although it was preliminary, it was largely classified into major categories 1, middle categories 2, and 3 levels, and documents that could occur in each field were proposed. As a result, it could be largely classified into corresponding documents, project documents, internal documents, and reference materials, and subsequent classifications could be derived. Documents that may occur in the decommissioning project must be managed by distinguishing between types to reduce the time for duplication or search, and the capacity of the storage can be efficiently managed. Therefore, it is hoped that the document types considered in this study will be used as reference materials for the decommissioning project and develop into a more systematic structure.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

During the operation of the nuclear power plant, various radioactive waste are generated. The spent resin, boron concentrates, and DAW are classified as a generic radioactive waste. They are treated and stored at radioactive waste building. In the reactor vessel, different types of radioactive waste are generated. Since the materials used in reactor core region exposed to high concentration of neutrons, they exhibit higher level of surface dose rate and specific activity. And they are usually stored in spent fuel pool with spent fuel. Various non-fuel radioactive wastes are stored in spent fuel pool, which are skeleton, control rod assembly, burnable neutron absorber, neutron source, in core detector, etc. The skeleton is composed of stainless 304 and Inconel-718. There are two types of control rod assembly, that are WH type and OPR type. The WH type control rod is composed of Ag-In-Cd composites. The OPR type control rod is composed of B4C and Inconel-625. In this paper, the characteristics and storage status of the non-fuel radioactive waste will be reported. Also, the management strategy for the various non-fuel radioactive waste will be discussed.

The segmentation of activated components is considered as a one of the most important processes in decommissioning. The activated components, such as reactor vessel and reactor vessel internals, are exposed to neutron from the nuclear fuel and classified to intermediate, low, and very low-level wastes. As it is expected, the components, which are closed to nuclear fuel, exhibit higher degree of specific activity. After the materials were exposed to neutrons, their original elements transform to other nuclides. The primary nuclides in activated stainless steel are 55Fe, 63,59Ni, 60Co, 54Mn, etc. The previous study indicates that the specific activity of individual nuclide is strongly depends on the material compositions and impurities of the original materials. The 59Co is the one of the most important impurities in stainless steel and carbon steel. In this paper, the relationship between individual nuclides in activation analysis of activated components was studied. The systematic study on specific activity of primary nuclides will be discussed in this paper to understand the activation tendency of the components.

Dry active waste (DAW) contains substantial amount of cellulose related materials. The DAW are usually classified as low and/or very low-level waste. In Korea, three types of disposal facilities have been considered: silo, engineering barrier, and land-fill. Currently, only the silo type disposal facility is in operation. Around 27 thousand drums were disposed in silo. Massive amount of cement concrete is used in construction of silo. The ground waste, which flow through the concrete structure, shows higher pH than as it is. It is generally known that the pH of silo is ~12.47 in Korea, when considering construction material, filling material, and property of ground water. It is expected that the cellulose in DAW will be partially transformed to isosaccharinic acid (ISA). It is generally accepted that the ISA plays a negative role in safety analysis of disposal facility by stimulation of specific nuclides. Various factors affect the degradation of cellulose containing radioactive waste, such as degree of polymerization, pH of disposal condition, interaction between concrete structure and ground water, etc. In this paper, the disposal safety analysis of cellulose containing radioactive, usually paper, cotton, wood, etc., are studied. The degradation of cellulose with respect to degree of polymerization, pH of neighboring water, filling material of silo, etc. are reviewed. Based on the review results, it is reasonable to conclude that the substantial amount of DAW could be disposed in silo.

Nuclear weapon generates huge amount of radioactive fallout which is extremely dangerous. The fallout gradually falls to the ground and then covers every surface in city and nature. A hydrogel decontamination medium has been developed to clean the surface polluted by the fallout. The hydrogel is soluble in water so the used hydrogel can be simply removed from the surface by washing. However, significant amount of waste water, containing the radioactive fallout, is generated with this process. In this respect, it is necessary to secure alternative technical options for the used hydrogel recovery. In this study, a steam-suction process was suggested for the used hydrogel recovery. Contaminated stainless steel surface, with fixed simulated fallout particles, was prepared for test. The simulated fallout particles were obtained by high-temperature treatment of a mixture of natural soil, used concrete, and Fe2O3. The hydrogel, composed of poly-vinyl alcohol and borax, was spread onto the contaminated stainless steel surface. The hydrogel was soft at first and it gradually becomes rigid with time. The used hydrogel was recovered by suction with a simultaneous steam spraying to soften the rigid gel. As a result, the clean surface of the stainless steel without the simulated fallout particles was obtained, showing the feasibility of this technique for the used hydrogel recovery.

The operation and decommissioning of nuclear power plants (NPPs) creates waste in the process of handling radioactively contaminated material, which must be disposed of in a repository or can be recovered of in the same way as conventional waste in the course of handling radioactively contaminated materials. For buildings or sites of NPPs it also has to be decided under what conditions they can continue to be used for other, conventional purposes or demolished. This decision is referred to as “release from supervision under nuclear and radiation protection law” or “clearance” in short. The clearance levels applicable in Germany according to the Radiation Protection Ordinance have been defined such that a radiation dose (hereinafter referred to as “dose”) of 10 μSv per year is not exceeded. The vast majority of the materials resulting from the dismantling of a nuclear power plant (e.g. most of the massive concrete structures) are neither contaminated nor activated. Thus, there is no need to treat these materials as radioactive waste. Emplacement of uncontaminated masses which in Germany is essentially several million tonnes of building rubble in a repository would require additional construction of such facilities, which, in view of the negligible hazard potential, from the point of view of the Nuclear Waste Management Commission (ESK) is clearly to be rejected both economically and, in particular, ecologically. Alternative ways are increasingly discussed in public, such as the abandonment of buildings after gutting, i.e. refraining from demolition of the controlled area buildings of NPPs. Also, another proposal discussed in public, the landfilling or the long-term storage of cleared material at the site, does not offer any safety-related advantages either in the view of the ESK. If, after completion of all dismantling work, the building has been decontaminated such that the clearance levels for buildings are complied with further use of the building rubble resulting from demolition is harmless from a radiological point of view. For these reasons, Germany has deliberately decided to use clearance as an essential measure in the dismantling of NPPs. If it is intended to conventionally reuse or depose of virtually contaminant-free material from controlled areas, it must first undergo a clearance procedure. The prerequisites that must be fulfilled for clearance are regulated in the Radiation Protection Ordinance, which includes two basic clearance pathways: unrestricted and specific clearance. In the following, the basic process of clearance is briefly presented and illustrated for a better understanding. It comprises five steps. Step 1-Radiological characterization by sampling, Step 2-Dismantling of plant components in the controlled area, Step 3- Decontamination, Step 4-Decission measurements, Step 5-Clearacnce and further management. The entire clearance process is governed by a clearance notice and is carried out under the supervision of the competent authority under nuclear and radiation protection law or the independent authorized expert commissioned by it. The clearance pathways contained in the Radiation Protection Ordinance have proven themselves in practice. They permit safe and proper management of material from dismantling and release of the site from supervision under nuclear and radiation protection law. These German regulatory procedures should be taken into account and deregulation and removal should be used as appropriate and necessary tools in the process of decommissioning NPPs in order to return non-hazardous materials to the material cycle or for conventional disposal.

The removal of cesium (Cs) from contaminated clay minerals is still a challenge due to the limited efficiency of the process. Thus, this study aimed to enhance the removal for Cs+ ions during the conventional acid washing process by incorporating a bead-type adsorbent. Polyacrylonitrile-based nickel potassium hexacyanoferrate (NiFC-PAN) was utilized as the Cs adsorbent to selectively adsorb Cs+ ions in a strongly acidic solution that contained competing ions. To enable easy separation of clay particles and protect the adsorbent from harsh environmental conditions, PAN was deliberately constructed as large beads. The synthesized adsorbent (NiFC/PAN in a 2:1 ratio) displayed high selectivity for Cs+ ions and had a maximum capacity of 162.78 mg/g for Cs+ adsorption in 0.5 M HNO3 solution. Since NiFC-PAN exhibited greater Cs selectivity than the clay mineral (hydrobiotite, HBT), adding NiFC-PAN during the acid washing substantially increased Cs desorption (73.3%) by preventing the re-adsorption for Cs+ ions on the HBT. The acid treatment in the presence of NiFCPAN also significantly decreased the radioactivity of 137Cs-HBT from 209 to 27 Bq/g, resulting in a desorption efficiency of 87.1%. Therefore, these findings suggest that the proposed technique is a potentially useful and effective method for decontaminating radioactive clay.