인터넷과 빅데이터 시대에 혼돈을 기반으로 한 디지털 영상암호화 알고리즘에 대한 연구는 정보보안 분야 의 연구 핫스팟 중 하나가 되었다. 기존의 텍스트 데이터에 비해 디지털 이미지는 데이터의 양이 많고 중복 성이 높은 특성을 가지고 있습니다. DES, AES 및 RSA와 같은 기존 방법은 디지털 이미지 암호화 요구 사 항을 충족할 수 없습니다. 초기 민감도와 고유한 임의성 때문에 카오스는 암호화의 혼란 및 확산 개념과 매 우 일치하므로 카오스 이미지 암호화에 대한 연구가 빠르게 발전했습니다. 본 논문에서는 주로 혼돈 영상 암호화에 대한 연구를 수행한다. 평문 영상에 대한 키 의존성을 목표로 비트 평면과 개선된 로지스틱 맵 기 반 영상 암호화 알고리즘을 제안한다. 알고리즘은 스크램블된 일반 텍스트 이미지의 낮은 4비트 비트 매트 릭스를 개선된 물류 맵의 제어 매개변수로 사용하여 일반 텍스트 이미지에 대한 키의 종속성을 높이고 놀 라운 일회성 비밀 특성을 갖습니다. 서로 다른 비트의 이미지 정보에 따라 서로 다른 강도의 암호화 알고리 즘을 설계하여 알고리즘의 효율성을 높입니다. 영상 암호화의 확산 모드를 목표로 하이퍼 카오스 시스템과 블록 모듈로 기반의 영상 암호화 알고리즘을 제안한다. 알고리즘은 하이퍼 카오스 시스템을 사용하여 카오 스 시퀀스를 반복적으로 생성하고, 암호화된 픽셀 주위에 임의의 이미지 블록을 분할하고, 블록의 픽셀 값 에 대해 적분 모듈로 연산을 수행하여 암호화된 픽셀 회색 값을 얻음으로써 확산 범위를 효과적으로 확장 하고 확산의 무작위성을 향상시킵니다.

The canine parvovirus (CPV) causes clinical signs, such as severe enteritis, dehydration, diarrhea, vomiting, leukopenia, and hair loss, which may lead to death. Vaccination is still the most important approach, as no specific treatment exists to prevent CPV. Monoclonal antibodies are valuable tools to study the pathogenic mechanisms of CPV and develop effective diagnostic reagents and pharmaceuticals. In this study, two monoclonal antibodies (MAbs) against CPV-2a were obtained through hybridoma technology by fusing myeloma cells and B cells from the spleens of mice immunized with CPV type 2a (CPV-2a). Two MAbs (CPV-330 and CPV-620) were studied on the reactivity of vaccine (CPV-2a) and field strains (CPV-new 2a, -2b, and -2c) by indirect immunofluorescence (IFA), hemagglutination inhibition test (HI), virus neutralization test (VN), and inhibition of virus growth test. Two MAbs showed similar antibody titers for HI and VN. On the other hand, CPV-330 inhibited the viral replication in Crandell-Rees Feline Kidney (CRFK) cells better than CPV-620. These CPV MAbs may provide valuable biological reagents to study the CPV pathogenic mechanisms and work as therapeutic antibodies.

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.

본 연구에서는 액체감쇠역전회복(FLAIR) 시퀀스를 대체하였던 방법 중에 비교적 간단하면서 높은 재현성을 나타내었던 고신호 강도제거 원리를 MAGiC에 적용하여 MAGiC-FLAIR와 기존의 고속스핀에코-FLAIR 영상과 비교하여 고신호 강 도제거영상의 유용성과 임상적으로 유의미한 기준을 제시하고자 하였다. 연구방법은 MAGiC 적용 후 MAGiC-FLAIR와 MAGiC-고신호 강도제거 영상을 재구성하여, 기존의 고속스핀에코-FLAIR 영상과 각각 정성적, 정량적 평가를 비교하였 다. 정성평가결과 MAGiC-고신호 강도제거는 MAGiC-FLAIR 보다는 월등히 우수하며, 고속스핀에코-FLAIR와 유사한 결과를 보였고, 정량평가결과 MAGiC-고신호강도제거는 MAGiC-FLAIR보다 백질과 회백질 대조도는 더 우수할 뿐만 아 니라, 뇌척수액의 신호 억제도 우수한 결과를 나타냈다. Synthetic 영상을 통하여 획득한 다양한 대조도 영상 중 FLAIR의 부정확도를 고신호 강도제거기법을 적용한다면 진단적 가치를 개선하여 제공할 수 있을 것이다.

본 연구는 의료분야에서 정확한 환자 확인(patient identification)을 위한 의료생체인식기술(medical biometrics)을 소개하고 자 한다. 첫째, 생체인식기술(biometrics)에 대한 정의, 분류 및 종류와 같은 기본 정보다. 둘째, 생체신호 (biological signal)와 의료영상(medical imaging)을 활용한 의료생체인식기술의 종류와 최근 연구에 관한 정보를 기술했 다. 마지막으로 의료 환경에서 적용되는 생체인식기술의 종류와 사례와 함께 의료 환경의 현주소를 언급했다. 환자확인이 라는 궁극적인 목적을 가진 의료생체인식기술은 의료 현장에 점진적으로 도입되리라 생각한다. 본 연구는 본 연구가 정확 한 환자확인을 위한 의료생체인식기술을 이해하고 연구방향에 도움이 되는 기초 자료로 활용될 것으로 사료된다.

To develop quality-improved muffins, the influence of the partial replacement of wheat flour with lemon balm powder (LBP) on their quality characteristics was analyzed. Studies were carried out to examine the supplementation of different percentages of LBP (2%, 4%, 6%, and 8%) on the quality characteristics of LBP muffins. The incorporation of LBP significantly affected the physicochemical parameters of muffins. Such incorporation at different levels significantly decreased pH, moisture content, baking loss, hardness, and color parameters, including L*, a*, and b* values of muffins (p<0.05). No significant effect of LBP substitution was found in height changes (p>0.05). Finally, the consumer acceptance test indicated that the highest levels of LBP incorporation (i.e., 8%) had a considerably adverse effect on consumer preferences in all attributes. In contrast, muffins with moderate levels of LBP (4%) showed a good and satisfactory sensorial acceptance in general. Thus, LBP was successfully employed in muffins, improving quality and broadening its potential applications in other bakery products.

Appropriateness of the minimum detectable activity in the analysis of gamma radionuclides is very important. This is reason determine the time factor among the conditions of the analysis when it is rationally determined has the advantage that radioactivity analysis can be performed accurately and quickly. In this study, 100 mL of an unknown sample was diluted in Marinelli Beaker 1L to obtain, review data on gamma radiation analysis results and minimum detectable activity for each measurement time. The measurement was used High Purity Germanium detector, target nuclides are Co-57, Co-58, Y-88 and Cs-137. Since the radioactivity analysis sample will be expected to be the waste subject to selfdisposal or less during the radioactive waste classification, the minimum detectable activity standard was set based on the detection of less than the permissible activity for self-disposal for each nuclide. The measurement methods were measured by classifying it into seven categories: 1000 seconds, 3600 seconds, 10000 seconds, 30000 seconds, 80000 seconds, 100000 seconds, and 150000 seconds. The radioactivity from this measurement are Co-57 2.89 Bq·g−1, Co-58 0.19 Bq·g−1, Y-88 0.20 Bq·g−1, Cs-137 0.15 Bq·g−1, the measurement results under all conditions were similar. On the other hand, the minimum detectable activity showed values above the allowable activity for self-disposal in not but Co-58 at 1000 and 3600 seconds. Only after taking the measurement time of 10000 seconds, the result was derived Co-57 0.0095 Bq·g−1, Co-58 0.0068 Bq·g−1, Y-88 0.0052 Bq·g−1, Cs-137 0.0062 Bq·g−1, which was confirmed to less than the allowable activity for self-disposal by nuclide. Reasonably determining the measurement time in gamma radionuclide analysis is a very important issue in terms of economy of time and accuracy of measurement. Although this study cannot be said to be able to determine a reasonable measurement time for all gamma radionuclide analysis, it is hoped that research on various samples will be made to contribute to the efficient measurement of gamma radioactivity.

Gamma-ray spectroscopy, which is an appropriate method to identify and quantify radionuclides, is widely utilized in radiological leakage monitoring of nuclear facilities, assay of radioactive wastes, and decontamination evaluation of post-processing such as decommissioning and remediation. For example, in the post-processing, it is conducted to verify the radioactivity level of the site before and after the work and decide to recycle or dispose the generated waste. For an accurate evaluation of gamma-ray emitting radionuclides, the measurement should be carried out near the region of interest on site, or a sample analysis should be performed in the laboratory. However, the region is inaccessible due to the safety-critical nature of nuclear facilities, and excessive radiation exposure to workers could be caused. In addition, in the case of subjects that may be contaminated inside such as pipe structures generated during decommissioning, surveying is usually done over the outside of them only, so the effectiveness of the result is limited. Thus, there is a need to develop a radiation measurement system that can be available in narrow space and can sense remotely with excellent performance. A liquid light guide (LLG), unlike typical optical fiber, is a light guide which has a liquid core. It has superior light transmissivity than any optical fiber and can be manufactured with a larger diameter. Additionally, it can deliver light with much greater intensity with very low attenuation along the length because there is no packing fraction and it has very high radiation resistant characteristics. Especially, thanks to the good transmissivity in UV-VIS wavelength, the LLG can well transmit the scintillation light signals from scintillators that have relatively short emission wavelengths, such as LaBr3:Ce and CeBr3. In this study, we developed a radiation sensor system based on a LLG for remote gamma-ray spectroscopy. We fabricated a radiation sensor with LaBr3:Ce scintillator and LLG, and acquired energy spectra of Cs-137 and Co-60 remotely. Furthermore, the results of gamma-ray spectroscopy using different lengths of LLG were compared with those obtained without LLG. Energy resolutions were estimated as 7.67%, 4.90%, and 4.81% at 662, 1,173, and 1,332 keV, respectively for 1 m long LLG, which shows similar values of a general NaI(Tl) scintillator. With 3 m long LLG, the energy resolutions were 7.92%, 5.48%, and 5.07% for 662, 1,173, and 1,332 keV gamma-rays, respectively.

In this study, the positions of Cs-137 gamma ray source are estimated from the plastic scintillating fiber bundle sensor with length of 5 m, using machine learning data analysis. Seven strands of plastic scintillating fibers are bundled by black shrink tube and two photomultiplier tubes are used as a gamma ray sensing and light measuring devices, respectively. The dose rate of Cs-137 used in this study is 6 μSv·h−1. For the machine learning modeling, Keras framework in a Python environment is used. The algorithm chosen to construct machine learning model is regression with 15,000 number of nodes in each hidden layer. The pulse-shaped signals measured by photomultiplier tubes are saved as discrete digits and each pulse data consists of 1,024 number of them. Measurements are conducted separately to create machine learning data used in training and test processes. Measurement times were different for obtaining training and test data which were 1 minute and 5 seconds, respectively. It is because sufficient number of data are needed in case of training data, while the measurement time of test data implies the actual measuring time. The machine learning model is designated to estimate the source positions using the information about time difference of the pulses which are created simultaneously by the interaction of gamma ray and plastic scintillating fiber sensor. To evaluate whether the double-trained machine learning model shows enhancement in accuracy of source position estimation, the reference model is constructed using training data with one-time learning process. The double-trained machine learning model is designed to construct first model and create a second training data using the training error and predetermined coefficient. The second training data are used to construct a final model. Both reference model and double-trained models constructed with different coefficients are evaluated with test data. The evaluation result shows that the average values calculated for all measured position in each model are different from 7.21 to 1.44 cm. As a result, by constructing the double-trained machine learning model, the final accuracy shows 80% of improvement ratio. Further study will be conducted to evaluate whether the double-trained machine learning model is applicable to other data obtained from measurement of gamma ray sources with different energy and set a methodology to find optimal coefficient.

The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

In this work, we introduce a 100 kW class mobile plasma melting system designed for non-combustible radioactive wastes treatment. To ensure mobility, the designed system consists of two 24-ft commercial containers, each in charge of the plasma utilities and melting process. In the container for plasma utilities, a 100 kW class DC power supply is installed together with a chiller and gas supply system whereas the container for melting process has a transferred type arc melter as well as off-gas treatment system consisting of a heat exchanger, filtrations, scrubber and NOx removal system. As a heat source for a transferred type arc melter, we adopted a hollow electrode plasma torch with reverse polarity discharge structure. Detailed design for a 100 kW class mobile plasma melting system will be presented together with the main specifications of the components. In addition, the basic performance data of the melting system is also presented and discussed.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

As the plan for the nuclear dismantlement due to the permanent shutdown of Kori-1 and Wolseong- 1 nuclear power plants has been concretized, a “movable radionuclide analysis system” is being developed that can quickly and accurately analyze large amounts of radioactive waste generated on the sites during dismantling. This system has various advantages from the perspective of strict regulations on the radioactive waste movement and social acceptability, such as preventing unexpected accidents while moving on the national highway or expressway, reducing various documents and immediate response to dismantling plans. Currently the system is being developed to be equipped with previously developed sample pretreatment and radioactivity measuring equipment and automated volatile and nonvolatile nuclide separation equipments, but to ensure mobile stability, it needs to analyze factors and establish stability standards. In the KS Q ISO/IEC 17025:2017 standard, the requirements for “facilities and environmental conditions” are a very important factor in building reliability for consumers as part of the quality guarantee for this facility. In order to meet the requirements, the technical standards of various test equipment to be installed in this facility were investigated. The physical, chemical, and radiological hazards that could affect the safety of the equipment and workers in the process of moving the equipment between nuclear power plants or between nuclear dismantling sites were derived from vibrations, rapid changes in temperature and humidity, and the spread of contamination from radioactive waste samples. Therefore, the scope of application of the law, which is the basis for securing stability during movement, was classified into two situations: movement from facility manufacturer to installation site (non-contaminated) and movement from primary to secondary use (contaminated). And in order to investigate the Nuclear Safety Act, enforcement ordinances, and radiation safety management, and to establish standards for packaging and transportation of radioactive materials, the results of transportation tests and transport details were compared and analyzed. Finally, the air suspension systems and the automatic temperature and humidity control devices were analyzed to establish standards for securing stability against the vibration and the sharp changes in the temperature and humidity, and countermeasures such as accident measures in accordance with the Enforcement Decree of the Nuclear Safety Act were also investigated.

With the development of the nuclear industry and the increase in the use of radioactive materials, the generation of radioactive waste is increasing. As the generation of radioactive waste increases, the occurrence of related safety accidents is also increasing, and it is necessary to develop a radioactive waste monitoring technology to prevent such accidents in advance and efficiently manage radioactive waste. In Information and Communication Technology (ICT), various ICT technologies such as Internet of Things (IoT), Augmented Reality (AR), and Virtual Reality (VR) that can help with the safety management of these radioactive wastes are being developed. In this study, a radioactive waste monitoring technology was developed using ICT technology, such as management of the entire cycle history of waste using Quick Response (QR) codes, and development of AR visualization technology for small packages of radioactive waste. In addition, by using IoT technology to collect desired data from sensors and store the results, after the waste drum is loaded in the waste storage, a technology was developed to track and monitor the history and movement of the waste drum from repackaging to transfer to the storage. The data required for monitoring the radioactive waste drum includes location information, whether the drum is open or closed, temperature and humidity, etc. To collect this information, a drum monitoring technology was built with a 2.4 G wireless router, an anchor constituting a virtual zone, a tag to be mounted on the drum container, and a WNT server that collects sensor data. The network tool provided by WirePas was used for network configuration, and the status of gateways and nodes can be monitored by interworking with the WNT server. The configured IoT sensor technology were tested in a waste storage environment. Four anchors were installed and linked to the network to match the virtual zone and the real storage zone, and it was confirmed whether the movement of the tag was recorded on the network while moving the tag including the IoT sensor for analyzing location information. Based on these research results, it can contribute to the safety management of radioactive waste and establishment of Waste Acceptance Criteria (WCP) by and managing the history and monitoring the waste in the entire cycle from repackaging to disposal.

Dry head end process is developing for pyro-processing at KAERI (Korea Atomic Energy Research Institute). Dry processes, which include disassembly, mechanical decladding, vol-oxidation, blending, compaction, and sintering shall be performed in advance as the head-end process of pyroprocessing. An important goal of the head-end process is the fabrication of a proper feed material for the subsequent electrolytic reduction process. In the vol-oxidation process, the pellet type-SFs are pulverized by an oxidation under an air-blowing condition, and some volatile fission products are removed from the produced powders by using an air flow. After blending, the U3O8 powders are moved to a compactor of compaction process to obtain U3O8 porous pellets. In the fine powders removal system connected with compactor, for the improved performance of oxide reduction process coupled to dry head-end process, the removal/recovery system for fine powders potentially attached to the surface of oxide reduction raw material was developed and applied to the removal of fine powders from green pellets fabricated in dry head-end process. The removal efficiency of fine powders was also verified using porous U3O8 pellets in the fine powders removal system.

For economic and safe management of Spent Nuclear Fuel (SNF), it is very important to maintain the structural integrity of SNF and to keep the fuel undamaged and handleable. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration. The structural evaluation of SNF is very complicated and numerically demanding and it is essential to develop a simplified model for the fuel rod. In this study, a simplified model was developed using a new cladding failure criterion. The simplified model was developed considering only the horizontal or lateral static load utilizing the cladding material properties of irradiated Zirclaoy-4, and applicability in horizontal and vertical drop impacts was investigated. When a fuel rod is subject to bending, a very complicated 3D stress state is generated within the vicinity of the pellet–pellet interface. A very localized stress concentration is observed in the area where the edges of the pellets contact the cladding. If the failure strain criteria obtained from the uniaxial tension test or biaxial tube test is applied, failure is predicted at the beginning stage of loading with premature through-thickness stress or strain development. The localized contact stress or strain is self-limiting and is not a good candidate for the cladding failure criteria. In this work, a new cladding failure criterion is proposed, which can account for the localized stress concentration and the through-thickness stress development. The failure of the cladding is determined by the membrane plus bending stress generated through the thickness of the cladding, which can be calculated by a process called stress linearization along the stress classification line. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through horizontal and vertical drop impact simulations. It is shown that the stress correction factor derived considering static bending loading can be effectively applied to the dynamic impact analyses in both horizontal and vertical orientations.

In this study, for thermal neutron absorption, an aluminum metal composite in which B4C particles were uniformly dispersed was prepared using stirring casting and hot rolling processes. The microstructure, thermal neutron absorption rate, mechanical properties and dispersibility of the reinforcement of the prepared B4C/Al composite were analyzed. The composite in which the 40 μm sized B4C particles were uniformly dispersed increased the tensile strength as the volume ratio of the reinforcement increased.

An accumulation of spent nuclear fuel (SNF) has brought a considerable interest due to its energy and environmental issue. To effectively manage SNF, a pyroprocessing is introduced to separate useful resources from the spent fuels and to manufacture suitable fuels. In head-end process of pyroprocessing, spent fuels are thermally treated to prepare UO2 pellets, where various radioactive gases from SNFs are released during thermal treatment. Within these gases, C-14 as CO2 form is a radioactive fission product which had a long half-life of 5,730 years and emits beta radiation of 0.156 MeV. Generally, current CO2 capturing technologies include adsorption by solid materials, absorption by aqueous solutions, and membrane separation. Among these methods, absorption is an effective approach which traps CO2 effectively and and it is easy to operate at room temperature. In addition, it is highly recommended as immobilizing 14CO2 as CaCO3 formation due to the high thermal and chemical stability, and the relatively low solubility in water. Generally, a double alkali method has been proposed to capture low concentrated 14CO2 from the stream. This method for CO2 capture includes absorption process with NaOH solution and causticization using Ca(OH)2. In this study, CO2 emitted from SNF is captured using double alkali method, and the effects of operating conditions on capturing efficiency were investigated. Furthermore, considering the two-film theory, the effects of trapping conditions on the CO2 absorption performance were examined. The recovered CaCO3 from causticization was collected from the absorbing solution and analyzed.