Combustion method has been widely used in the analysis of 3H and 14C in various types of radioactive wastes since X. Hou reported the analysis of 3H and 14C in graphite and concrete for decommissioning of nuclear reactor. In this work, it was demonstrated that the validation result of combustion method for the simultaneous analysis of 3H and 14C in non-combustible radioactive wastes. To validate the combustion method, 3H and 14C spiked to sea sand and soil were prepared and each simulated sample was combusted with the accordance to a combustion temperature program. The validation of combustion method was assessed by the radioactivity recovery, radioactivity deviation, and relative standard deviation of measured radioactivity. The results of radioactivity recovery, radioactivity deviation, and relative standard deviation of 14C were 100~105%, less than 7%, less than 5%, respectively. In addition, 3H showed about 90% of radioactivity recovery, less than 20% of radioactivity deviation, and less than 5% of relative standard deviation. It will be provided that the validation results of combustion method in detail.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

Corrosion products generated from the oxidation of structure materials are deposited on the surface of coolant systems, forming CRUD (Corrosion Related Unidentified Deposits). The CRUD deposition on the fuel surface has influenced the heat transfer through the fuel rod. When CRUD was deposited on a fuel surface, heat resistance may increase, and this increase in heat resistance leads to the increase in temperature distribution from cladding to coolant. Also, the temperature distribution is related to the radiolytic and chemical reactions within the CRUD deposits. This influence may be enough to change the pH distribution within the CRUD deposits. To estimate the influence of thermal resistance, the composition, microstructure, and vapor fraction within the CRUD should be considered, by investigating the thermal conductivity model of CRUD deposits. Therefore, in this study, the CRUD thermal conductivity was studied through the literature study, by considering composition, capillary flow characteristics, and vapor fraction. For the uncertainty parameters, a sensitivity study was conducted to check the degree of influence on thermal conductivity. The effective thermal conductivity was applied to the radiochemistry model within the CRUD deposits and an analysis of the influence in radiolysis reaction within the CRUD deposits with a fixed thickness.

Molten chloride salts are promising candidates as a coolant for Molten Salt Reactors (MSRs) because of their low cost, high specific heat transfer, and thermal energy storage capacity. The NaCl- MgCl2 eutectic salts have enormous latent heat (430 kJ/kg) and financial advantage over other types of molten chloride salt. Despite the promise of the NaCl-MgCl2 eutectic salt, problems associated with structural material corrosion in the MSR system remain. The hygroscopicity of NaCl-MgCl2 and high MSRs operating temperature accelerate corrosion within structural alloys. Especially, MgCl2 reacts with H2O in the eutectic salt to produce HCl and Cl2, which are known to further exacerbate corrosion by the chlorination of structural materials. Therefore, several studies have worked to purify impurities associated with MgCl2, such as H2O. Thermal salt purification of NaCl-MgCl2 eutectic salt is one method that reduces HCl and Cl2 gas generation. However, MgO and MgOHCl are generated as the byproduct of thermal purification through a reaction between MgCl2 and H2O. The corrosion behavior of MgO within structural alloys after thermal treatment is not well known. This paper demonstrates corrosion behavior within structural alloy after thermal treatment at various temperature profiles of the NaCl-MgCl2 eutectic salt. According to the temperature range, MgCl2·H2O are separated at 100~200°C, and MgOHCl and HCl begin to occur at 240°C or higher. Finally, MgOHCl produces MgO and HCl at 500°C or higher temperatures. After thermal treatments, the H2O, MgOHCl, and MgO content were measured by Thermo Gravimetric Analyzer (TGA) to evaluate significant products causing corrosion. The structural materials were analyzed by the Scanning Electron Microscope-Energy Dispersive Spectroscopy (SEM-EDS) and using the mass change method to observe the type of localized corrosion, the corrosion rate, and the corrosion layer thickness. This study is possible in that it can reduce economic costs by reducing the essential use of expensive, high-purity molten salts because it is related to a substantial financial cost problem considering the amount of molten salt used in industrial sites.

Molten salts based on magnesium chloride can be used in the nuclear power reactor because they have a high heat capacity and heat stability, and allow for a faster neutron spectrum. However, magnesium chloride is highly hygroscopic, leading to the formation of moisture-related impurities, which result in the corrosion of structural materials and negatively affect the operation of the reactor. The dehydration of magnesium chloride is studied using both thermal and electrochemical treatments. According to previous studies, water impurities in magnesium chloride molten salt transform into magnesium oxide over 650 degrees Celsius. The temperature profile of the molten salt is suggested to separate magnesium chloride and magnesium oxide, focusing on cooling rate near the freezing point of magnesium chloride. Two layers separated by a phase boundary on the salt surface appear due to the density difference between magnesium oxide and magnesium chloride. Further, the removal of oxide ions remaining in the molten salt is carried out by electrochemical treatment. Two different cells, each consisting of two electrodes, are used. One cell is composed of graphite anode and nickel cathode. The other is composed of tin oxide anode and nickel cathode. As the reaction proceeds, carbon dioxide and oxygen are generated in graphite and tin oxide, respectively, and magnesium electrodeposition occurs at the cathode. The amount of purified magnesium oxide is measured to the endpoint, which is notified by the reduced current. The efficiency of each method is compared by measuring the weight ratio of the purified part to the unpurified part. Thermogravimetric analyzer (TGA) and UV-vis spectroscopy are used to check the quality of the purified part. Only magnesium oxide remains at a temperature above the boiling point of magnesium chloride. Therefore, the amount of magnesium oxide in the purified part can be measured by the mass change of the salt through the TGA method. For UV-vis spectroscopy, the transmittance is measured which depends on the weight percent of the impurities in the purified part. The suggested purification method using both thermal and electrochemical treatment is assessed quantitatively and qualitatively. It is expected that hygroscopic molten salts other than magnesium chloride will be able to be dehydrated through the above process.

In this study, molten salt experiments were performed using a multi-purpose molten salt experimental loop to evaluate the corrosion and thermodynamic properties of the molten salt. The multi-purpose molten salt experimental loop is made of 1-inch austenitic 316 stainless steel, and 1/4-inch austenitic 316 stainless steel tubes were welded on the surface of a 1-inch pipe to measure temperatures of molten salt. During the experiment, the molten salt leaked due to corrosion of the welded part of the 1/4-inch tube connected to the 1-inch pipe. Therefore, the cause of corrosion damage of the leaked welded part was analyzed. The effect of NaCl-MgCl2 salt selected as the molten salt on corrosion failure was considered. And based on the operation data of molten salt experiments, the time of occurrence of the issue was estimated. Lastly, the cause of corrosion failure was estimated by comparing and analyzing the pipe shape before and after failure using SEM-EDS.

Cardiovascular disease remains as one of the most common causes of high morbidity and mortality worldwide, despite remarkable medical advances in recent decades. Non-invasive techniques play a preeminent role in prevention of cardiovascular disease by diagnosing it at an early stage and guiding optimal patient management. Nuclear imaging is one of the most powerful means available for noninvasive diagnosis and management of poorly perfused myocardial region resulting from the cardiovascular disease. Several radionuclides are available for monitoring blood flow to cardiac tissue. The most validated radionuclides for these measurements are 13N, 15O, 99mTc, 201Tl and 82Rb. Each of 13N, 15O and 201Tl require the presence of an on-site cyclotron, whereas, 82Rb and 99mTc require only a generator. Rubidium (Rb) is an alkali metal ion that acts biologically like potassium and accumulates in cardiac muscle tissue. Rb has a rapid blood clearance profile which allows the use of 82Rb. It also has an ultra-short physical half-life of 75 sec for non-invasive evaluation of regional cardiac blood flow. There are several advantages of 82Rb over other radionuclides. Having a short half-life significantly reduces the radiation dose to the patient. In addition, 82Rb is a positron emitter, which gives the full advantages of PET such as image quantification with superior sensitivity. Several reports have shown superior diagnostic performances of 82Rb-PET over conventional 99mTc-SPECT. 82Rb can be produced from a generator system by the decay of its 25.6-day half-life parent 82Sr. However, the 82Sr parent is difficult to prepare. In routine generator production, certain purity is required to meet the specification of the product. Since there has been no the use of 82Rb radionuclide for research or medical purpose in Korea, we have plans to produce 82Sr with certain purity and develop a 82Sr/82Rb generator system. These studies can also be applied to remove radioactive Sr from radioactive waste waters. Because ion exchange resin, used for purification of 82Sr from impurities, is also utilized to trap radioactive Sr2+ ions from radioactive waste water. After Fukushima Daiichi nuclear accident, interest in the treatment of radioactive waste water has surged. As one of main fission products of nuclear reactor, 90Sr has been regarded as a hazardous radionuclide with half-life of about 29 years. Therefore, the investigation on ion exchange resin is important for removal of 90Sr from radioactive waste water. Here, we optimized 82Sr purification method using ion exchange resin to establish the most suitable procedure.

Attempts to use the molten salt system in various aspects such as MSR or energy storage systems are increasing. However, there are limitations in the molten salt-assisted technique due to the harsh corrosiveness of the molten salt, and a more detailed study on salt-induced corrosion is needed to solve this problem. In this study, corrosion behaviors of 80Ni-20Cr alloy in various salt environments such as eutectic NaCl-MgCl2 with NiCl2, CrCl2, and EuCl3 additives were investigated. Meanwhile, the corrosion acceleration effects of 80Ni-20Cr specimens were analyzed for various ceramic materials such as SiC, Al2O3, SiO2, graphite, and BN, and metallic materials such as Ni-based alloy, Fe-based alloy, and pure metals in a molten salt environment. The experiments were conducted at 973 K for up to 28 days, and after the experiment, the microstructural change of the specimen was analyzed through SEM-EDS, and salt condition was analyzed by ICP-OES.

With respect to the geologic repository, intrusion of groundwater has been considered as a major factor that can transfer radionuclides to the natural environment. Moreover, the migration of radionuclides in the natural groundwater system is significantly influenced by the interaction between the radionuclides and groundwater constituents. Among various hydrogeochemical reactions, hydrolysis is one of the major reactions that can affect the aqueous solubility of radionuclides. Therefore, a precise understanding of relevant chemical thermodynamic behavior is of cardinal importance for the reliable prediction of migration/retardation behavior of radionuclides in the natural groundwater system. The objective of the present work is to investigate the solubility behavior of Nd(OH)3(s) to provide relevant chemical thermodynamic data of Nd(III) as a chemical analogy of major radiotoxic elements such as Am(III) and Cm(III). All the experiments were performed with Ar gas-filled glovebox under inert atmospheric condition. The aqueous Nd(III) solution was prepared by dissolution of 0.5 g NdCl3·6H2O (Sigma-Aldrich) in 10 ml of deionized water. The Nd(III) solid phase was precipitated by dropwise addition of ca. 10 ml of 4 M NaOH (Sigma-Aldrich). The Nd(III) precipitate was identified to be crystalline Nd(OH)3(s) nanorod by using XRD and TEM. For the solubility experiment, the solid Nd(OH)3(s) was equilibrated at the pH range from 5.0 to 9.0 at 0.1 M NaCl condition. The total concentration of the Nd(III) was quantified by using UV/Vis absorption spectroscopy and ICP-MS after the phase separation. In the present work, the solubility behavior of the solid Nd(OH)3(s) phase was investigated by using colorimetric analysis. The chemical thermodynamic data obtained in this study are expected to enhance the reliability of solubility prediction for the trivalent lanthanides and actinides.

The massive amount of radioactive waste will generated during decommissioning of nuclear. Among the radioactive waste from these disposal process, 50-55 million tons of concrete waste are included. For safe disposal, it is very important to accurately analyze the concentration of radionuclides, especially 129I and 131I, contaminated concrete. 129I, a long-lived radioisotope of iodine (t1/2=1.57 × 107 y), and 131I (t1/2=8.04 d) are generated from the fission of uranium in nuclear reactors. In Korea, according to the Nuclear Safety and Security Commission (NSSC) radioactive clearance level guide, the limit for radioactive clearance level of 129I is less than 0.01 (Bq/g). Iodine can be absorbed, accumulate in organisms, and exhibit low energy emission compared with cesium, and cobalt. Therefore, it is essential to an accurately separate and analyze iodine radioactive waste. In this study, we focused on the determination of iodine in simulated cement waste form containing KI for the recovery of iodine. We performed cement waste form sieved through a different particle size (0.5 mm < ɸ < 6.35 mm). For the separation of iodine from solid samples with low iodine content, such as soil, sediment, and cement, for sample decomposition associated with solvent extraction using CHCl3 for separation of iodine from the matrix. The separation of iodine in cement waste particles was therefore carried out using an acid leaching method using KI containing cement particles. We observed that cement particle size decreased at 6.35 mm to 0.5 mm with iodine yield decrease at 0.840±0.011 to 0.582±0.010. Thus, in this study, the acid leaching method enables to determination Iodine in cement.

We conducted multi-elements determination of reference material certified by the Inorganic Ventures, IV-26, using iCAP 7400 ICP-OES of Thermo Fisher Scientific. And we statistically evaluated analysis results by introducing the in-house proficiency evaluation method implemented at the Ministry of Food and Drug Safety. Ca, Co, Fe, Mg, Ni, and V were selected as target elements, and extended uncertainty was estimated at a confidence level of about 95% and coverage factor k = 2. Five parameters incurred at manufacturing process (standard solution, calibration curve, repeated measurement and dilution factor of the test sample) were considered when determining the uncertainty. En-score can be calculated using the formula En=(x-X)/(Ulab 2+Uref 2)1/2 described in KS Q ISO 13528, where x, Ulab, X, and Uref are the test results, the uncertainty of the result, and the certified value and the uncertainty of the value. And if the absolute value |En| is less than 1, it can be evaluated as a satisfied value. As a result of ICP-OES analysis, each concentration of the elements to be measured was almost similar to the certified concentration of the reference material, and the uncertainty was slightly different. Also since evaluation on multi-elements determination had an En-score within 1, it was confirmed that the analysis results satisfied En evaluation.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

Neptunium (Np) is one of the daughter elements included in the decay chain of Pu. The quantitative analysis of Np isotopes is required for radioactive waste characterization, research on actinide chemistry, etc. Np-237 has a long half-life (2.144 million years), but its daughter Pa-233 has a relatively short half-life (26.975 days). For this reason, after a sufficient time elapses following the chemical preparation process of the analyte, the two nuclides are in radiation equilibrium in the sample. Np-237 emits alpha-rays while Pa-233 emits beta-rays. Both nuclides also emit gamma- and X-rays. In this study, alpha-rays were measured using liquid scintillation counting (LSC) method and alpha spectrometry. Gamma-spectrometry with a HPGe detector was used for the analysis of gammaand X-rays. In addition, we compared the radiometric results with quantitative analysis of Np using UV-Vis absorption spectrometry. The LSC method and the HPGe gamma-spectroscopy do not require extensive sample preparation procedures. Alpha spectroscopy requires a standard material spiking, separation by coprecipitation, and disk-type sample preparation procedure to obtain measurement efficiency and recovery factor. A reference material sample with a concentration of 5.8 mM was analyzed by the four analysis methods, and all of the measured results agreed well within a difference level of 4%.

Complexing agents used at various nuclear facilities exist in low- and intermediate level radioactive wastes deposited in the repository site. In addition these will be generated through the degradation of the wastes such as cellulose materials. The presence of chelating agents may possibly affect the safety of the wastes repository by promoting the migration of radionuclides into geosphere. Thus, under Nuclear Safety and Security Commission’s Notice No. 2021-16, the contents of chelating agents in radioactive wastes are required to be determined for the secured disposal. UV-Vis method based on an enzymatic reaction was proved to be in adequate to analyze citric acid in radioactive wastes with complex matrix, especially for concrete. A rapid automated method using ion chromatography (IC) for analysis of citric aicd in concrete samples is developed. This automated method enables a sample solution to measure without pretreatment and a target substance to separate from other concrete admixtures. Also, the developed method here, for radioactive concrete wastes was successfully applied to real samples with lowering a limit of quantification value.

Molten salt reactor (MSR) is one of the non-pressurized-water fourth-generation reactors that uses liquid nuclear fuel that integrates coolant and nuclear fuel, so it is a safe reactor that can fundamentally prevent severe accidents caused by coolant loss. MSR uses NaCl-MgCl2 as a coolant salt, which is considered a promising diluent that can dissolve the fuel salt by forming an eutectic mixture. In this study, a zone-melting system was used to remove impurities from the NaCl-MgCl2 used in MSR. The system was designed in detail to control eutectic salt impurities by traversing long charges into a small molten zone.

Anderson-type polyoxometalate (POM) with general formula of [Hy(XO6)M6O18]n- (y=0-6, n=2-8, M=addenda atom, X=heteroatom) represents one of the basic topological structures among POM-type family. Anderson-type POMs have a planar arrangement and two terminal oxygen atoms attached to each addenda metal atom unlike other type. Thus, the Anderson-type POMs have high reactivity and various coordination modes. The various multifunctional organic-inorganic hybrid materials can be synthesized using the Anderson-type POMs as an inorganic building block. Another important feature of the Anderson-type POMs is the incorporation of the heteroatoms with various sizes and oxidation states, which can lead to tune chemical properties. No Anderson-type POMs with early transition metal ions in the heteroatom site have been reported previously. Recently, we reported the synthesis of titanium-containing Anderson-type POM, Na2K6Ti0.92W6.08O24∙12H2O (Ti-POM), which consists of pure inorganic framework built from a central Ti atom surrounded by six WO6 inorganic scaffold. Herein, in-depth studies were conducted to find optimal synthesis conditions such as composition and pH. The success of synthesis was confirmed with Powder X-ray Diffraction that the Ti-POM has a rhombohedral structure with space group of R-3m (No. 166) when the TiOSO4·xH2SO4∙yH2O/ Na2WO4∙2H2O molar ratio is in the range of 0.07 to 0.33. But outside of this range, other unwanted phases coexist. In a basic condition (pH=12), a single-phase Ti-POM with good crystallinity can be obtained, while a Keggin-type POM, NaxK10-x(H2W12O42), was formed through the decomposition of Ti-POM at pH lower than 7.

In Korea, 483,102 assemblies of spent fuel have been discharged and stored in sites, as of 2019. However, total capacity for site storage is 529,748 assemblies, and more than 90% is already saturated. Wolsong site, the most saturated site, started to construct more dry storage to extend the capacity in 2020. Spent fuel and high-level waste (HLW) is a big concern in Korean nuclear industry. Then, master plan for management of spent fuel is once announced by Ministry of Trade, Industry and Energy (MOTIE) in 2016 and reviewed by civil committee in 2019. The core contents of the plan are establishing schedule for construction of HLW management facility in one area, and construction of temporary dry storage in each site, if unavoidable. For HLW management facility, there are three following schedules: siting of Underground Research Laboratory (URL) and Interim Storage by 2020, operation of facilities initiated by 2030, and operation of final disposal facility initiated by 2050. Final repository will be designed as deep geological repository. The concept of deep geological disposal is that spent nuclear fuel is placed in disposal containers that can withstand corrosion and pressure in long-term, permanently isolated from the human sphere of life, and dumped in deep geological media, such as crystalline rocks and clay layer, at a depth of 300 to 1,000 meters underground. The safety assessment of waste disposal sites focuses on determining whether the disposal sites meet the safety requirements of national regulatory authority. This safety assessment evaluates the potential radiation dose of radionuclides from the disposal site to humans or the environment. In this case, the calculation is performed assuming that all engineering barriers of the disposal site have collapsed in a long-term period. Then radionuclides are released from the waste, and migrated in groundwater. The dose resulting from the release and migration of radionuclides on the concentration of nuclides in groundwater. In general, metallic nuclides may exist in water in various ionic states, but some form colloids. This colloid allows more nuclides to exist in water than in solubility. Therefore, more doses may occur than we know generally predict. To determine the impact of colloids, we performed the safety assessment of the Yucca Mountain repository as an example.

Viscosity is a fundamental physical property that is important in any system in which fluid movement occurs. In addition, most of the elements exist as ions in molten state in high-temperature molten salt, and electrical conductivity in such molten state is closely related to viscosity as a transport property. Molten salt reactor (MSR) and pyroprocess are representative processes dealing with high-temperature molten salts, actinide elements, and other radioactive materials. In MSR and pyroprocesses, the viscosity data must be provided as one of the fundamental physical property data required for safe process operations and countermeasures to severe accidents. In order to measure the viscosity of highly corrosive molten salt at high temperatures, we have built a in-house developed molten salt viscosity measurement system based on the Brookfield rotationary viscometer. We also developed a special correction technique to improve the accuracy of the viscosity measurement. In this study, the viscosity was measured at 500°C for NaCl-MgCl2 molten salt, which is selected as the base salt material of MSR system under development in Korea Atomic Energy Research Institute (KAERI), using our viscosity measurement system installed in a oxygen- and moisture-free Ar-atmosphere glovebox. Our viscosity measurement system was calibrated using a LiCl-KCl eutectic mixture with well-known viscosity value, and viscosity values obtained using our own correction methodology were compared with those of other conventional correction methods. In our further study, we plan to measure the NaCl-MgCl2-UCl3 system at various compositions and temperatures.

Hydrogen-bonded organic frameworks (HOFs) are a new type of porous crystalline material that are constructed by intermolecular hydrogen bonding of organic building blocks to form twodimensional (2D) and three-dimensional (3D) crystalline networks. High-quality HOF single crystals are easily grown for direct superstructure analysis using single crystal X-ray diffraction, which is essential for revealing the relationship between structure and properties. The unique advantages of HOF, such as high crystallinity, porosity and fast regeneration, have allowed it to be used in a variety of applications including catalysis and gas separation. Squaric acid (SQA) is a non-carboxylic, organic acid with proton donor and acceptor ability which is known to take on a variety of coordination modes with metal ions. Pyrazine is a six-membered aromatic heterocycle bearing two nitrogen atoms, which has sp2 hybridized C atoms with C-H hydrogen bonds. This work describes the synthesis and structural characteristics of HOF based on squaric acid and pyrazine. Based on single crystal X-ray diffraction data, this MOF crystallizes in the triclinic P-1 space group. Each asymmetric unit is composed of H2SQ and pyrazine. All squaric acid molecules share one H atom with the N atom of pyrazine molecules. The layer distance between nearby O atoms from squaric acid in different layers equals 5.29 Å. Also, our HOF showed high adsorption capacity the during experiments. The composition and comparative characteristics of HOF are given using SCXRD, PXRD, SEM and UV-vis.