When decommissioning of nuclear facilities happens, large amounts of radioactive wastes are released. Because costs of nuclear decommissioning are enormous, effective and economical decontamination technologies are needed to remove radioactive wastes. During NPP operation, corrosion product called Chalk River Unidentified Deposits (CRUD) is generated. CRUD is an accumulation of substances and corrosion products consisting of dissolved ions or solid particles such as Ni, Fe, and Co on the surface of the NPP fuel rod coating. CRUD is slowly eroded by the circulation of hot pressurized water and later deposits on the fuel rod cladding or external housing, thereby reducing heat production by the nuclear fuel. Decontamination of radiologically contaminated metals must be performed before disposal, and several methods for decontaminating CRUD are being studied in many countries. Decontamination technology is an alternative to reducing human body covering and reducing radioactive waste disposal costs, and much research and development has been conducted to date. Currently, the importance of decontamination is emerging as the amount of waste stored in radioactive waste storage is close to saturation, and the amount of radioactive waste generated must be minimized through active decontamination. In this study, a preliminary study was conducted on the removal of CRUD by multiple membrane in an electro-kinetic process using an electrochemicalbased decontamination method. Preliminary research to develop a technology to electrochemically remove CRUD by using a self-produced electrochemical cell to check the pH change over time of the CRUD cell according to voltage, electrolyte, membrane and pH change.

The treatment process for Spent Filter(SF) of Kori-1 was developed that includes the following : 1) Taking out by robot system 2) Screening by ISOCS 3) Collection of representative samples using a sampling machine 4) Compression 5) Immobilization 6) Packaging and nuclide analysis and 7) Delivery/disposal. Although the robot system, ISOCS, sampling machine and immobilization facility are essentially required for building the above processing but decision to build the compression system and nuclide analysis system must be made after reviewing the need and cost benefit for their construction. In addition, for effcient SF treatment, it is necessary to determine the nuclide concentration range of the SF to which immobilization will be applied. In this study, a cost benefit analysis was performed on existing and alternative methods for processes related to compression treatment, nuclide analysis and immobilization methods, which are greatly affected by economics and efficiency according to the design. First, although the disposal cost is reduced with reducing the number of packaging drums by compressed and packaged but the expected benefits not be equal to or greater than the cost invested in building a compression system. As a result, non-compressed treatment of SF is expected to be economical because the construction cost of compression system is more expensive than the benefits of reducing disposal costs by compression. Second, a cost benefit analysis of direct and indirect nuclide analysis methods was performed. For indirect analysis, scaling factors should be developed and the drum scanner suitable for the analysis for DAW should be improved. As a result, direct analysis applied grouping options is expected to be more economical than indirect analysis requiring the cost for developing scaling factors and improving the scanner. Third, it is timeconsuming and inefficient to distinguish and collect filters that are subject to be immobilized according to the waste acceptance criteria among the disorderly stored SFs in the filter rooms. If the benefits of immobilization of the SFs selectively are not greater than the benefits of immobilization of all SFs, it can be economical to immobilize all SFs regardless of the nuclide concentration of them. As a result, it is more economical to immobilize all SFs with various nuclide concentrations than to selectively immobilize them. The conclusion of this study is that it is not only cost-effective but also disposal-effective to design the treatment process of SF to adopt non-compressed processing, direct analysis and immobilization of all SFs.

Decommissioning waste is generated with various types and large quantities within a short period. Concrete, a significant building material for nuclear facilities, is one of the largest decommissioning wastes, which is mixed with aggregate, sand, and cement with water by the relevant mixing ratio. Recently, the proposed treatment method for volume reduction of radioactive concrete waste was proven up to scale-up testing using unit equipment, which involved sequentially thermomechanical and chemical treatment. According to studies, the aggregate as non-radioactive material is separated from cement components with contaminated radionuclides as less than clearance criteria, so the volume of radioactive concrete waste is decreased effectively. However, some supplementation points were presented to commercialize the process. Hence, the process requires efficiency as possible to minimize the interface parts, either by integration or rearranging the equipment. In this study, feasibility testing was performed using integrated heating and grinding equipment, to supplement the possible issue of generated powder and dust during the process. Previously, heat treatment and grinding devices were configured separately for pilot-scale testing. But some problems such as leakage and pipe blockage occurred during the transportation of generated fine powder, which caused difficulties in maintaining the equipment. For that reason, we studied to reduce the interface between the equipment by integrating and rearranging the equipment. To evaluate the thermal grinding performance, the fraction of coarse and concrete fines based on 1mm particle size was measured, and the amount of residual cement in each part was analyzed by wet analysis using 4M hydrochloric acid. The result was compared with previous studies and the thermomechanical equipment could be selected to enhance the process. Therefore, it is expected that the equipment for commercialization could be optimized and composed the process compactly by this study.

When aluminum is in an alkaline state, the aluminum oxide film surrounding aluminum is dissolved and moisture penetrates the exposed aluminum surface, causing corrosion of aluminum. At this time, hydrogen gas is generated and there is a risk of explosion due to the generated hydrogen gas. Aluminum radioactive waste is difficult to permanently dispose of because it does not meet the standards for the acquisition of low- and intermediate-level radioactive waste cave disposal facilities currently managed and operated by the Korea Nuclear Environment Corporation. However, because of this risk, it is necessary to study how to safely treat and dispose aluminum waste. In this study, overseas cases were investigated and analyzed to ensure the safety of aluminum waste disposal, and the current status of aluminum radioactive waste generated during decommissioning of the Korea Research Reactor 1&2 and a treatment plan to secure disposal suitability were presented. The process of removing a little remaining oxygen in molten steel during the reduction of iron oxide in the iron refining process is called deoxidation, and a representative material used for deoxidation is aluminum. In the case of metal melting decontamination, which is one of the decontamination processes of radioactive metal waste, a method of treating aluminum waste by using aluminum as a deoxidizer is proposed.

Wet solid wastes including spent ion exchange resins, evaporator concentrates and sludges require solidification to transform wastes into an acceptable solid, monolithic form for final disposal. The development of the process control program for the solidification of radioactive sludges generated at nuclear power plants has been in progress to provide reasonable assurance that the solidified product will meet the established waste acceptance criteria for solidified waste. A mobile solidification system to produce the solidified waste in the size of a 200 L drum was used, which adopts the in-line mixing method where the waste and binder are mixed and then transferred to the disposable container. To simulate radioactive sludges, non-radioactive sludges are synthesized and the specimens are prepared by using them. The qualification tests on the prepared specimens including the compressive strength test, the thermal cycling test, the irradiation test, the leach test, the immersion test, etc. have been performed to qualify recipes for a range of waste compositions. The results of the tests will be analyzed and discussed.

Dry active wastes (DAWs) are combustible waste generated during the operation and decommissioning of nuclear facilities, and are known to be generated in the amount of approximately 10,000 to 40,000 drums (based on 200 L) per unit. It consists of various types of protective clothing, paper, and plastic bags, and is stored in radioactive waste storage facilities. Therefore, reducing the volume of DAWs is an important issue in order to reduce storage costs and utilize the limited space of waste storage facilities. Heat treatment such as incineration can dramatically reduce the volume of waste, but as the waste is thermally decomposed, CO2, a global warming gas, is generated and there is a risk of emissions of harmful gases including radionuclides. Therefore, a heat treatment process that minimizes the generation of CO2 and harmful gases is necessary. One of the alternatives to incineration is to carbonize DAWs, dispose of carbonized materials below the release standard as non-radioactive waste, and selectively separate and stabilize inorganic components, including radionuclides, from carbonized DAWs. In this study, 13 types of DAWs generated from nuclear power plants were selected and their thermal decomposition characteristics were investigated to design a heat treatment process that replaces incineration. As a result of TGA analysis, the temperature at which thermal decomposition of each waste begins is 260-300°C for cotton, 320-330°C for paper, 315-420°C for synthetic fiber, 350°C for latex gloves. The mass of most samples decreased to less than 1 % of the initial weight after heat treatment, and dust suit and latex gloves had residues of 13.83% and 13.71% of the initial mass, respectively. The metal components of the residue produced after heat treatment of the sample were analyzed by EDS. According to the EDS results, cotton contains Ca and Al, paper contains Ca, Al and Si, synthetic fiber contains Ca, Cu and Ti, latex gloves contain Ca and Mg. Additionally, ICP analysis was performed to quantify the inorganic components. These results are expected to be applicable to the processing of DAW generated at nuclear facilities in the future.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

There is a large amount of radioactive waste in waste storage in the Korea Atomic Energy Research Institute. Some of the radioactive waste was generated during the dismantling process due to Korea Research Reactor 1&2 and it accounts for 20% of the total waste. Radioactive waste must be reduced by appropriate disposal methods to secure storage space and to reduce disposal costs. Research Reactor wastes include wastes that are below the acceptable criteria for selfdisposal and non-contaminated wastes, so they can be treated as wastes subject to self-disposal through contamination analysis and reclassification. In order to deregulation radioactive waste, it is necessary to meet the self-disposal standards stipulated in the Domestic Nuclear Act and the treatment standards of the Waste Management Act. The main factors of deregulation are surface contaminant, radionuclide activity and dose assessment. To confirm the contamination of waste, surface contaminant and gamma nuclide analysis were performed. After homogenizing the waste sample, it was placed in 1 L Mariinelli beaker. When collecting waste samples, 1 kg per 200 kg of waste was collected. The concentrations of the major radionuclides Co-60, Cs-134, Cs-137, Eu-152, and Eu-154 were analyzed using HPGe detector. To evaluate radiation dose, various computational programs were used. A dose assessment was performed with the analyzed nuclide concentration. The concentrations of representative nuclides satisfied the deregulation acceptance criteria and the results of the dose assessment corresponding to self-disposal method was also satisfied. Based on this results, KAERI submitted the report on waste self-disposal plan to obtain approval. After final approval, Research Reactor waste is to be incinerated and incineration ash is to be buried in the designated place. Some metallic waste has been recycled. In this study, the suitability of deregulation for self-disposal was confirmed through the evaluation of the surface contaminant analysis, radionuclide concentration analysis and dose assessment.

The effectiveness of a crystalline natural barrier in providing sealing capabilities is based on the behavior of numerous fractures and their intersections within the rock mass. It is important to evaluate the evolving characteristics of fractured rock, as the hydro-mechanical coupled processes occurring through these fractures play a dominant role. KAERI is actively developing a true tri-axial compression test system and concurrently conducting hydro-mechanical experiments using replicated fractured rock samples. This research is focused on a comprehensive examination of coupled processes within fractures, with a particular emphasis on the development of true tri-axial testing equipment. The designed test system has the capability to account for three-dimensional stress conditions, including vertical and both maximum and minimum horizontal principal stresses, realizing the disposal conditions at specific underground depths. Notably, the KAERI-designed test system employs the mixed true tri-axial concept, also known as the Mogi-type, which allows for fluid flow into fractures under tri-axial compression conditions. This system utilizes a hydraulic chamber to maintain constant stress in one direction through the application of oil pressure, while the other two directional stresses are applied using rigid platens with varying magnitudes. Once these mechanical stress conditions are established, control over fluid flow is achieved through the rigid platens in contact with the specimen section. This pioneering approach effectively replicates in-situ mechanical conditions while concurrently observing the internal fluid flow patterns within fractures, thereby enhancing our capacity to study these coupled phenomena. As future research, numerical modeling efforts will be proceeding with experimental data-driven approaches to simulate the coupled behavior within the fractures. In these numerical studies, two distinct fracture geometry domains will be generated, one employing simplified rough-walled fractures and the other utilizing mismatched rough-walled fractures. These investigations mark the preliminary steps in the process of selecting and validating an appropriate numerical model for understanding the hydro-mechanical evolution within fractures.

Spent nuclear fuels (SNFs) are stored in nuclear power plants for a certain period of time and then transported to an interim storage facility. After that, SNFs are finally repackaged in a disposal canister at an encapsulation plant for final disposal. Finland and Sweden have already completed the design of the spent nuclear fuel encapsulation plant. In particular, Finland has begun the construction of the encapsulation plant and is on the verge of completion. Korea Radioactive Waste Agency (KORAD) is conducting a conceptual design of a deep geological repository for SNFs. Conceptual design of the encapsulation plant is part of the research activity. It is highly required to draft an operation process of the encapsulation plant before an actual design activity. As part of the activity, Finnish design concept of the encapsulation plant and experience were thoroughly reviewed. Finally a preliminary concept of the operation process was proposed considering Korean unique situations such as the volume of SNFs estimated to be disposed of, types of transportation cask and other considerations.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

The process basket assembly is an important module in pyroprocess, because pyroprocess is a batch process, so process materials are contained in a basket assembly and transferred with the basket. The basket assembly is composed of upper and lower assembly. The lower assembly is a basket or crucible which contains process materials, and it can have electrodes. The upper assembly mainly consists of heat shields, a flange, and connectors for supplying currents to electrodes of the lower assembly. During the electrolytic recovery process, the lower part is submerged into molten salt, whose temperature is about 500°C at least and the heat from salt is transferred to the upper assembly. And the heat affects the performance or durability of parts on the top of equipment and can raise cell temperature, which is an undesired situation. In addition, the handling equipment can pick the assembly when it is hottest, and during the transfer, the gripped part is under thermal and mechanical stresses. Because of this, the thermal effects from the heat should be required during equipment design stage. In this study, the thermal analyses of process basket assembly were conducted for 3 cases: the steady state of the basket assembly when it submerged in molten salt, the thermal and mechanical stresses when gripped by remote handling device, and the temperature changes under natural convection. These analyses were performed using Solidworks with flow simulation package, and the results will apply to improve the thermal resistant performance of the basket assembly.

Pyroprocessing technology has emerged as a viable alternative for the treatment of metal/oxide used fuel within the nuclear fuel cycle. This innovative approach involves an oxide reduction process wherein spent fuel in oxide form is placed within a cathode basket immersed in a molten LiCl-Li2O salt operating at 923 K. The chemical reduction of these oxide materials into their metallic counterparts occurs through a reaction with Li metal, which is electrochemically deposited onto the cathode. However, during process, the generation of Li2O within the fuel basket is inevitable, and due to the limited reduction efficiency, a significant portion of rare earth oxides (REOx) remains in their oxide state. The presence of these impurities, specifically Li2O and REOx, necessitates their transfer into the electrorefining system, leading to several challenges. Both Li2O and REOx exhibit reactivity with UCl3, the primary electrolyte within the electrorefining system, causing a continuous reduction in UCl3 concentration throughout the process. Furthermore, the formation of fine UO2 powder within the salt system, resulting from chemical reactions, poses a potential long-term operational and safety concern within the electrorefining process.Various techniques have been developed to address the issue of UO2 fine particle removal from the salt, utilizing both chemical and mechanical methods. However, it is crucial that these methods do not interfere with the core pyroprocessing procedure. This study aims to investigate the impact of Li2O and REOx introduced from the electrolytic reduction process on the electrorefining system. Additionally, we propose a method to effectively eliminate the generated UO2 fine powder, thereby enhancing the long-term operational stability of the electrorefining process. The efficiency of this proposed solution in removing oxidized powder has been confirmed through laboratory-scale testing, and we will provide a comprehensive discussion of the detailed results.

The radionuclide management process is a conditioning technology to reduce the burden of spent fuel management, and refers to a process that can separate and recover radionuclides having similar properties from spent fuels. In particular, through the radionuclide management process, high heat- emitting, high mobility, and high toxicity radionuclides, which have a significant impact on the performance of disposal system, are separated and managed. The performance of disposal system is closely related to properties (decay heat and radioactivity) of radioactive wastes from the radionuclide management process, and the properties are directly linked to the radionuclide separation ratio that determines the composition of radionuclides in waste flow. The Korea Atomic Energy Research Institute have derived process flow diagrams for six candidates for the radionuclide management process, weighing on feasibility among various process options that can be considered. In addition, the GoldSim model has been established to calculate the mass and properties of waste from each unit process of the radionuclides management process and to observe their time variations. In this study, the candidates for the radionuclide management process are evaluated based on the waste mass and properties by using the GoldSim model, and sensitivity analysis changing the separation ratio are performed. And the effect of changes in the separation ratio for highly sensitive radionuclides on waste management strategy is analyzed. In particular, the separation ratio for high heat-emitting radionuclides determines the period of long-term decay storage.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

Various disposal methods for spent nuclear fuels (SNFs) are being researched, and one of these methods involves separating high heat-generating nuclear isotopes such as Strontium-90 (90Sr) and Cesium-137 (137Cs) for deep disposal. These isotopes has relatively short half-lives and substantial decay energies. Especially, 90Sr undergoes decay through Yttrium-90 to Zirconium-90, emitting intense heat with beta radiation. Therefore, the removal of these high heat-generating isotopes will significantly contribute to reducing disposal site area. To remove 90Sr from SNFs, molten salt was utilized in KAERI. During this process, it was discovered that 90Sr dissolves in the molten salt in the form of SrCl2 and/or Sr4OCl6. Afterwards, it is crucial to recover 90Sr in the form of oxide from the salt to create immobilized forms for disposal. This can be achieved by reactive distillation with K2CO3. However, the amount of 90Sr within the SNFs is only 0.121wt%, and even if all the 90Sr in the SNFs were to leach into the molten salt, the quantity of 90Sr in the molten slat would still be very small. Therefore, adding K2CO3 to the molten salt for reactive distillation could result in significant possibilities of side reactions occurring. In this study, a two-step process was employed to mitigate the side reactions: the 1st step involves evaporating the all molten salts and the 2nd step includes adding K2CO3 to make oxides through solid-solid reaction. Eutectic LiCl-KCl, which is the most commonly used salt, was employed. The eutectic LiCl-KCl with SrCl2 was heated at 850°C for 2 h to evaporate the salts under a vacuum (> 0.02 torr). However, after examining the distillation product before the solid-solid reaction, it was observed that SrCl2 reacted with KCl in the salt, resulting in the formation of KSr2Cl5. It means that salts containing KCl are not suitable candidates for reactive distillation aimed at producing immobilized forms. As an alternative, MgCl2 could be a highly promising candidate because it is inert to SrCl2 and according to a recent study in KAERI, MgCl2 exhibited the most efficient separation of Sr among various salts. Therefore, we plan to proceed with the two-step reactive distillation using MgCl2 for the future work.

This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.