1970년대 이후 게임시장 규모의 폭발적 성장과 함께 하드웨어의 발전을 바탕으로 온라인 게임의 성장을 비 롯하여 게임 규모는 지속적으로 확장하고 있다. 대규모 자본과 인력이 투입된 이른바 AAA게임은 게임의 시장규모를 증가시킨 순기능 이외에 게임플레이의 획일화를 가져온 역기능으로도 작용하게 되었다. 이러한 시장의 반대측면에서 소규모 또는 1인 개발자들에 의해 개발되는 인디게임이 차츰 주목받고 성장하고 있 는 추세에 있다. 특히 ‘마인크래프트’의 성공 이후에 인디게임 업계로 진입하는 소규모 게임 개발자와 1인 게임 개발자들의 양도 동반 증가하였으며, 이들은 AAA게임이 가지지 못한 독특한 아이디어와 디자인 요소 를 활용하고, 크라우드 펀딩 등의 방법을 통해 자금을 조달하는 제작 형태를 띄고 있다. 인디게임은 2022년 현재 게임 플랫폼 ‘스팀’에 등록된 게임수가 40043개에 이를 정도로 지속적인 시장 확대를 이루고 있으나 최근, 스폰서와 대규모 투자를 바탕으로 한 인디게임 개발의 추세에 힘입어 소규모 또는 1인 인디게임 개 발자들의 경쟁력은 낮아지고 있다. 본 연구에서는 2020년부터 2022년까지 ‘인디게임‘ 키워드를 바탕으로 빅 데이터 분석방법을 통하여 인디게임 산업의 이슈를 추출하고 이슈와 관련한 게임의 현황과 향후 인디게임 방향성을 밝혀보고자 하였다. 연구의 결과로 2020년의 인디게임 플랫폼의 보급과 함께 시장이 확장되는 것 을 알 수 있으며, 2021년은 인디게임 보급을 뒷받침 할 수 있는 지원정책이 확산되는 현상을 파악할 수 있 었다. 2022년은 다양한 해외 플랫폼을 바탕으로 지원정책에 힘입어 지속적으로 제작되고 있으며 해외진출 을 시도함과 동시에 인디게임 산업의 활성화를 위한 육성정책이 이루어지고 있음을 알 수 있다.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.

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.

Currently, treatment and disposal suitability verification methods have not been established for radioactive waste, such as spent filters temporarily stored in each plant, so the WCP (Waste Certification Program) can be applied to verify the suitability of non-conforming waste at the site. In this study, WCP components such as certification organizations, certification methods, certification documents, and quality assurance (QA) plan that should be considered when developing WCP applicable to spent filter disposal were reviewed and presented. First, a certification organization consists of a certification organization that performs certification work, a certification support organization related to waste generation and treatment, and a quality control organization for waste certification. Especially, the support organization should support the implementation of WCP, so that spent filter processing procedures such as generation information management and immobilization can be properly packaged and transported. Second, in identifying the waste characteristics of the certification method, each characteristic identification procedure and certification method of the acceptance criteria should be described, evidence examining the suitability of general, radiological, physical, chemical, and biological requirements, and processes related to measurement and sampling should be established. In identifying characteristics, satisfaction of waste form, free water requirements, and whether it is subject to immobilization should be checked priorly, and a method of confirming particulate matter and securing filling rate when packaging compressed filters should be included. It is very important to develop a technology for verifying the safety and quality of the immobilized material because immobilization of the filters can be a processing method that satisfies various characteristic criteria. Meanwhile, it is essential to collect samples and develop scaling factors to identify the nuclides of filters and prove that they are below the concentration limits. For chemical and biological requirements, the characteristics are identified through generation information documents, corrective actions are taken and documented in case of nonconformance. Third, certification documents should include immobilization procedure manual, characteristic report, and characteristic test manuals such as free water, particulate matter and filling rate, radiation measurement method manual for packages, profile, and generation documents. Fourth, the QA plan should analyze the QA system of the plants, check the QA inspection details, establish general requirements for QA of spent filter disposal, and specify step-by-step certification work QA activities. In this study, considerations to ensure the disposal suitability at all stages from generation to disposal of spent filter were presented, and development of a WCP could contribute to preventing nonconformance.

As the decommissioning of Kori Unit 1 progresses, securing technology for treatment and disposal of radioactive wastes that have not been disposed of so far, such as spent filters, is recognized as an urgent task. In this study, a method of confirming the disposal suitability of spent filters was presented by reviewing the waste characteristics as presented in the waste acceptance criteria (WAC). The waste characteristics to be satisfied to ensure disposal suitability of waste are largely classified into general requirements, solidification and immobilization requirements, radiological requirements, physical requirements, chemical requirements, and biological requirements. First, the general requirement is to prove that the prohibited waste form has not been introduced into items related to waste form and packaging, and to confirm the suitability of disposal through step-by-step packaging photos, generation information, X-ray inspection, and visual inspection. Second, in the solidification and immobilization requirements, spent filters are non-homogeneous waste, and if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1 or more, they must be immobilized. Third, in order to meet the characteristic criteria for nuclides and radioactivity concentration, sampling and scaling factors development are required and based on this, nuclides must be identified and demonstrated to be below the disposal concentration limits. Surface dose rate and surface contamination should be measured in accordance with standardized procedures and disposal suitability should be confirmed through document tests recording the measured values. Fourth, in order to satisfy the physical requirements of the particulate matter and filling rate characteristics, the spent filter must be immobilized, if necessary, thereby ensuring disposal suitability. Meanwhile, free water in the spent filter should be removed through pre-drying and dehydration, and the disposal suitability should be confirmed by applying a test. Fifth, the criteria for chelating agents should be checked for disposal suitability through operation records and component analysis of spent filters, and documents, that can prove harmful substances are removed in advance and no harmful substances are included in the package, should be provided. Lastly, in biological requirements, if the spent filters contain corrosive or infectious substances, they should be removed in advance and disposal suitability should be confirmed by providing documents that can prove that such substances are not included in the package.

This study established a process to ensure the disposal suitability of spent filters stored in the untreated state in Kori unit 1 and presented the following procedures and requirements for confirming the disposal suitability for each process. The process for securing spent filter disposal suitability consists of collecting spent filters, compression, immobilization, analysis and packaging, and storage stages. The requirements for confirming the acceptance criteria for each process are as follows. (1) Collecting: Since the high radioactivity spent filters are being stored in the filter room of Kori unit 1, those are collected by a remote system to minimize the exposure dose of workers due to spent filter handling. In order to satisfy the surface dose rate requirements, spent filters with a surface dose rate of 10 mSv·hr−1 or more are classified and collected, and stored temporary storage place until a separate treatment plan is determined. The checkpoints in this process are the surface dose rate, etc. (2) Compression: The collected spent filters are analyzed gamma nuclides such as Co-60 and Cs-137, using a field-applicable nuclide analyzer, and then applying the scaling factors to determine whether it is disposable. Spent filters whose radioactivity concentration is confirmed to be less than the disposal concentration limit is compressed into compression ratios determined by surface dose rate. The checkpoints in this process are nuclide information, surface dose rate, compression ratio, spent filter loading quantity, etc. (3) Immobilization: A spent filter is a non-homogeneous waste that is immobilized with a proven safety material such as cement if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1. Meanwhile, immobilization of inhomogeneous waste can be considered to satisfy disposal criteria such as particulate matter and filling rate. The checkpoints in this process are the immobilizing material, filling rate, etc. (4) Analysis and Packaging: Immobilized drums shall be determined to be 95% or more of the total radioactivity of waste packages by measuring the radioactivity concentration of nuclides using a nuclide analysis device. Finally, measure the surface dose rate and surface contamination of the package, and attach the package label recording the identification number, date, total radioactivity, surface dose rate, and surface contamination information to the packaging container. (5) Storage: Packaging containers are moved to and stored in a temporary waste storage or storage area before disposal.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.

FTIR (Fourier Transform Infrared) and Raman spectra of KJ-II bentonite provided by Clariant Korea were compared with those of MX-80 bentonite. The FTIR spectra were obtained using a Nicolet 5 FTIR spectrometer (Fisher Scientific) equipped with a diamond ATR (Attenuated Total Reflection) module. The spectra were collected for 64 scans with a resolution of 4 cm−1. Raman spectra were obtained using an optical microscope (Olympus, BX43) and a spectrometer (Andor, SR- 500). The laser beam was focused using an objective lens with a magnifying power of 50. The backscattered light from the sample was collected into an optical fiber with a core diameter of 0.4 mm. The Raman signals were recorded with CCDs (Andor, DV401A-BV for 532 nm laser wavelength and DV420A-OE for 638 and 785 nm laser wavelengths). Each pixel of CCD received the signal for 1 s and its 1000 times accumulated data were collected. The FTIR spectra of the two bentonite samples are very similar. The FTIR spectra of both bentonites showed absorption bands at 3623, 3399, 3231 cm−1 in the higher wavenumber region and at 1637, 1443, 1117, 997, 914, 887, 847, 797, 611, 515, 414 cm−1 in the lower wavenumber region. A sharp band at 3623 cm−1 and the strong band at 997 cm−1 correspond to the OH stretching of structural hydroxyl groups and the Si-O stretching vibration, respectively. In addition to these clear bands, several absorption bands observed in this experiment are well matched with the results reported in various literatures. Unlike the FTIR spectrum, it is not easy to observe the Raman bands of bentonite. The reason is that strong fluorescence interferes with the Raman spectrum. The two bentonite samples showed different fluorescence intensities. In the case of MX-80 bentonite, no clear Raman band was observed due to the influence of very strong fluorescence. KJ-II bentonite showed a relatively weak fluorescence intensity and Raman bands were partially visible at around 147, 260, 397, 709, and 1279 cm−1. In particular, the band at 1279 cm−1 is wide and sturdy. It was observed that the non-powder samples showed a better quality spectra. The Raman characteristics of KJ-II bentonite, which depend on the incident laser wavelength and the sample pretreatment, are discussed in detail.

Deep geological repository (DGR) has been considered as a globally accepted strategy to dispose high-level radioactive wastes. During long storage periods of 100,000 years, uranium (U) could be migrated through fractures in deep granite aquifers and interact with indigenous bacteria under anaerobic condition. Anaerobic bacteria can reduce U(VI) and further precipitate in the form of U(IV)-oxide minerals by transferring electrons through c-type cytochrome. In this point of view, a comprehensive understanding of uranium-microorganisms interaction is necessary to guarantee the safety of high-level radioactive waste disposal. Although diverse bacterial communities are present in DGR environment, a number of studies have been focused on some model bacteria, such as Desulfovibrio, Geobacter, and Shewanella spp.. In this study, indigenous bacterial community in deep granitic groundwater at 234–244 m was inoculated to sterile uranium-contaminated granitic groundwater amended with 20 mM of sodium acetate, and then incubated under anaerobic condition for 12 weeks. Bio-reduction of U(VI) to U(IV) by indigenous bacteria in uranium-contaminated groundwater was investigated during whole operation period. Initial U(VI) concentration of 885.4 μg·L−1 gradually decreased to 586.1 μg·L−1, resulting in approximately 33.8% of aqueous U(VI) removal efficiency. Oxidation-reduction potential (ORP) value was gradually decreased from 175.4 mV to –243.0 mV after the incubation of 12 weeks. The decrease in ORP value was attributed to the presence of aerobic bacteria and facultative anaerobic bacteria in indigenous bacterial community. The shift in bacterial community structure was observed by 16S rRNA gene high-throughput sequencing analysis. Proteobacteria (55.6%), Firmicutes (24.1%), Actinobacteria (5.5%), and Bacteroidetes (5.4%) were dominant in initial indigenous bacterial community, while Proteobacteria (94.8%) was found to be the only abundant phylum after the reaction. In addition, great increase in the relative abundance of sulfate-reducing bacteria (SRB) was observed: the relative abundance of SRB increased from 11.4% to 44.3% after the reaction. This result indicates that the SRB played a key role in the removal of aqueous U(VI). This finding shows the potential of aqueous U(VI) removal by indigenous bacteria in DGR environment.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, transporting the SNF from SFP to interim storage facilities or permanent disposal facilities should be considered. There are two ways to transport the SNF from a site to other site, one is the land transportation with truck or train, and the other is the maritime transportation with ship. The maritime transportation has some advantages compared with the land transportation. The maritime transportation method uses safer route which is far from populated area than land transportation method, and transport more weight than land transportation method. However, the cask should be loaded into the ship for the maritime transportation, and there is a possibility of a drop accident of the cask onto the ship. Therefore, it is necessary to evaluate the structural integrity of the cask and ship for the drop accident during the loading process. To evaluate the structural integrity of the cask and ship, it is necessary to determine the analysis conditions that caused the greatest damage in the drop accident. There may be various conditions such as the drop angle of the cask, the initial falling speed, the drop position onto the ship, the size of the ship, etc. This study set the drop angle of the cask and the drop position onto the ship as the simulation variables, which have high possibility to occur during cask drop. However, the others are excluded since they are controllable by worker. In this paper, various drop angle (0, 15, 30, 45, and 70 degree) of the cask were simulated to define the greatest damage condition. KORAD-21 cask model was used for Finite Element Analysis (FEA), and FEA was performed to simulate a horizontal drop (1 m drop). The strain-hardening material properties for the deck were used as HT36 steel. The Cowper-Symonds constitutive model for HT36 was used to consider the strain rate effect. A Tie-down structure for supporting the cask was modeled with the cask model which contained inner structures like canister, basket, etc. Structural integrity of the cask and tie-down structure were evaluated using the von-Mises stress and equivalent plastic strain (PEEQ), and one of the ship deck was evaluated using deflection of ship deck and equivalent plastic strain. Compared with each cask drop angle conditions, 45 degree of the cask drop angle showed the highest deflection and PEEQ values, but did not exceed ultimate strain of HT36. In the ship deck, the corner of deck showed the highest PEEQ value in all simulation cases. As the result, the 45 degree of the cask drop angle condition results was more conservative than other conditions, and the corners of deck failure was able to evaluate ship safety.