To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

Various types of radioactive liquid and solid wastes are generated during the operation and decommissioning of nuclear power plants. To remove radionuclides Co-60, Cs-137 etc. from a liquid waste, the ion-exchange process based on organic resins has been commonly used for the operation of nuclear facilities. Due to the considerations for the final disposal of process endproduct, other treatment methods such as adsorption, precipitation using some inorganic materials have been suggested to prepare for large amounts of waste during decommissioning. This study evaluated sintering characteristics for radioactive precipitates generated during the liquid waste treatment process. The volume reduction efficiency and compressive strength of sintered pellets were the major parameters for the evaluation. Major components of a simulated precipitate were some coagulated (oxy) hydroxides containing light elements, such as Si, Al, Mg, Ca, and zeolite particles. Green pellets compressed to around 100 MPa were heated at a range of 750~850°C to synthesize sintered pellets. It was observed that the volume reduction percentages were higher than 50% in the appropriate sintering conditions. The volume reduction was caused by the reduction of void space between particles, which is an evidence of partial glassification and ceramization of the precipitates. This result can also be attributed to conversion reactions of zeolite particles into other minerals. The compressive strength ranged from 6 to 19 MPa. These results also showed a significant correlation with the volume reduction of sintered body. Although our lab-scale experiments showed many benefits of sintering for the precipitates, optimized conditions are needed for large-scale practical applications. Evaluation of sintering characteristics as a function of pellet size and further testing will be conducted in the future.

KORAD (Korea Radioactive Waste Agency, http://www.korad.or.kr) has stored slightly contaminated ascon (asphalt coated concrete mixture) that was introduced to Gyeongju repository about a decade ago waiting for a final disposal. It is believed to be mainly contaminated by radioisotope 137Cs due to impurities introduced from the outside during the ascon manufacturing process. We studied characteristics of the radioactive waste to see whether this material would be proper enough to be disposed in Gyeongju LILW repository or be other ways to reduce the disposal volume including self-disposal before its final disposal otherwise. KORAD looked into the properness of characteristics of ascon in terms of WAC (Waste Acceptance Criteria) documented by KORAD that includes general chemical and physical properties of asphalt, density, size of grains, content of organic material and possibility of existence of chelate materials that qualitatively limited to be disposed by the criteria. And other associated characteristics such as gas generation and bio degradation were also investigated. Based on the data obtained from the study, we proposed various plausible solutions in associated with operational and disposal safety and economic view points. This study will be used for KORAD’s decision on how to control and safely dispose the spent ascon within a reasonable time period. And also those experiences may be applied for other LILW issues that require treatment or conditioning of radioactive wastes in the future.

Various types of solidifying materials are used to stabilize and solidify low and intermediatelevel radioactive dispersible waste. Portland cement is generally used to solidify various radioactive wastes because its facilities and processes are simple, less dangerous, and it has excellent compressive strength after curing compared to other materials. However, it is difficult to use Portland cement in radioactive waste containing highly water-soluble harmful substances such as sodium fluoride because it is prone to leaching harmful ingredients in immersion tests due to its low water resistance. In this study, solidification was achieved using an organic-inorganic hybrid solidifying binders consisting of inorganic binders such as Portland cement, blast furnace slag powder, silica fume, and organic binders such as epoxy resin. This material was then compared with a solidification material made of Portland cement alone. The mixing ratio of inorganic binders, water, and organic binders to simulated waste is 35%, 20%, and 25%, respectively. The mixing ratio of inorganic binders and water when using only Portland cement for simulated waste is 100% and 80%, respectively. The mixed paste was poured into a cylinder mold (Φ 5 × 10 cm) to seal the upper part, cured at room temperature for 28 days to produce a solidification specimen, and then subjected to various tests were performed, including compressive strength, immersion compressive strength, hydration peak temperature, length change, and immersion weight change. The compressive strength of the organic-inorganic hybrid solidification test was 13-17 MPa, the immersion compressive strength was 15-18 MPa, the hydration peak temperature was 33-36°C, the length change rate was -0.086%, and the immersion weight change rate was –2.359%. The compressive strength of the Inorganic solidification test using only Portland cement was 16-18 MPa, the immersion compressive strength was 20-21 MPa, the hydration peak temperature was 23-25°C, the length change rate was -0.150%, and the immersion weight change rate was -5.213%. The compressive strength and immersion compressive strength of the organic-inorganic hybrid solidification materials were slightly lower compared to those of Portland cement solidification materials, they still met the compressive strength standard of 7-12 MPa, taking into consideration the strength reduce and economic feasibility of the core drill process. Furthermore, it indicates that the rates of change in length and immersion weight decreased to about 1% and 5%, suggesting an improvement in water resistance. The above results suggest that applying the organic-inorganic hybrid solidification method to radioactive waste treatment can effectively improve water resistance and help secure long-term stability.

Radioactive liquid waste generated during the operation of domestic nuclear power plants is treated through a somewhat different liquid radwaste system (LRS) for each plant. Prior to the introduction of standard nuclear power plants, LRS used a concentrated water dry system (CWDS) to evaporate liquid waste and manage it in the form of dry powder. The boron-containing radioactive liquid waste dry powder was solidified using paraffin from 1995 to 2010, and about 3,650 drums (based on 200 L) of paraffin solidified drums are currently stored in nuclear power plants. Paraffin solidification drums do not meet the acceptance criteria for radioactive waste repositories because it is difficult to secure the homogeneity of the solidified body and there are concerns about leaching of radioactive waste due to the low melting point of paraffin. In order to solve this problem and safely permanently dispose of paraffin solidification drums, the characteristics of dry powder paraffin solidification drums containing boron-containing radioactive liquid waste must be analyzed and appropriate treatment technology utilizing the results must be introduced. This study analyzes the physical properties of paraffin, the chemical properties of boron-containing radioactive waste dry powder, and the physicochemical properties of paraffin solidification powder, and proposes an appropriate alternative technology for treating boron-containing radioactive waste dry drum. When disposing of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder, the solidification body must be effectively withdrawn from the drum and the paraffin must be completely separated from the solidification body. When disposing the drum, the solidified material must be effectively extracted from the drum and the paraffin must be completely separated from the solidified material. Afterwards, the paraffin must be self-disposed, and the radioactive waste must be disposed of in accordance with acceptance criteria of repository. We looked at how each characteristic of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder can be utilized in each of the above treatment processes.

Radionuclides in low- and intermediate-level radioactive wastes from the decommissioning process of nuclear power plants were generally immobilized by cementation methods. Ethylenediaminetetraacetic acid (EDTA), which is extensively used as a decontamination agent, can affect the behaviors of radionuclides immobilized in cement waste forms. In this study, the effects of EDTA contained in simulated radioactive decommissioning wastes on the leaching characteristics of immobilized Co and Cs and the microstructure evolution of cement waste form. Co leaching was accelerated by the formation of Co–EDTA complexes with high mobility and solubility. Cs leaching was hindered by the ion competition with other metal–EDTA complexes for releasing from the cement waste form. Cs leaching was also retarded by carbonated layer at edge of the cement waste form, which process is facilitated by the presence of EDTA. Finally, the effects of EDTA on the leaching characteristics of immobilized Cs and Co and the microstructure evolution of the cement waste form should be considered to ensure the safety of disposal for lowand intermediate-level radioactive wastes.

아메리카동애등에(H. illucens)는 음식물 폐기물 등 유기성 폐자원을 효율적으로 처리할 수 있는 능력을 가지 고 있어 전세계적으로 주목받고 있는 환경정화 곤충이다. 하지만 유기성 폐자원을 처리 시 가장 큰 문제는 아메리 카동애등에가 먹이인 유기성 폐자원을 소화시킬 때 발생되는 악취이다. 국내에서 현재 아메리카동애등에를 사육하고 있는 농가는 223호로 조사되고 있지만 이중 악취발생 저감장치 등을 설치한 농가는 10%가 안되는 것으 로 생각된다. 따라서 국내에서 동애등에 먹이로 가장 많이 사용되는 습식사료를 먹이로 사용하였을 때 농가 사육 장 안에서 발생되는 복합악취와 지정악취 22종에 대하여 분석하였다. 그 결과, 복합악취는 249배였으며, 지정악 취는 22종 중 7종(암모니아, 메틸메르캅탄, 트라이메틸아민, 아세트알데하이드, 프로피온알데하이드, 뷰틸알 데하이드, i-발레르알데하이드)가 검출되었다. 이중 가장 높은 농도를 나타낸 악취물질은 암모니아로 98.4ppm 이 분석되었다. 또한, 아메리카동애등에를 사육 시 가장 많이 발생되는 암모니아의 발생시기는 사육초기인 1~4 령보다 5령 이후 전생육기 중의 대부분을 발생시키는 것으로 조사되었다. 이러한 결과는 암모니아 저감을 위한 적정시기를 설정하는데 도움이 될 것으로 생각된다.

In this study, the growth characteristics of Lentinula edodes were confirmed by bean sprout waste(BW) as an alternative raw material for rice bran. The mycelium growth of Sanjo701, a major cultivation variety of L. edodes, was compared between a medium mixed with 8:2(v/v) of oak sawdust and a medium mixed with BW 50% and BW 100%. The mycelium growth in BW 50% was 13.5 cm. Compared to the control, BW 50% increased the diameter of the pileus by 1.6 cm. Additionally, the length of the pileus decreased by 0.4 cm when comparing the growth of the fruit body. In contrast, at BW 50%, the diameter of the pileus decreased by 9.6 cm and the length of the stipe decreased by 1.4 cm. According to analysis of the constituent amino acids, BW 50% showed a lower overall nutritional content than the control, whereas BW 100% had a lower amino acid content than the control. However, glutamic acid and aspartic acid, which are flavor-enhancing ingredients, were observed at levels of 3.954 mg/g and 1.436 mg/g, respectively, in BW 100%. Therefore, if bean sprout by-products are efficiently processed and utilized, it is believed that they will be beneficial to farmers as a substitute for rice bran and reduce the cost of manufacturing substrate

In this study, we investigate the opportunity of using waste tire char as a cathode material for lithium-ion primary batteries (LPBs). The char obtained by carbonizing waste tires was washed with acid and thermally fluorinated to produce CFX. The structural and chemical properties of the char and CFX were analyzed to evaluate the effect of thermal fluorination. The carbon structure of the char was increasingly converted to CFX structure as the fluorination temperature increased. In addition, the manufactured CFX- based LPBs were evaluated through electrochemical analysis. The discharge capacity of the CFX reached a maximum of 800 mAh/g, which is comparable to that of CFX- based LPBs manufactured from other carbon sources. On the basis of these results, the use of waste tire char-based CFX as a cathode material for LPBs is presented as a new opportunity in the field of waste tire recycling.

A large amount of small and medium-sized metal waste is generated during the decommissioning of nuclear power plants (NPPs). Metal waste is mostly contaminated with low-level radioactive, so it needs decontamination for self-disposal and recycling. A large amount of Organic Decontamination Liquid Waste during decontamination will be generated. The generated organic liquid waste is low in concentration, so the decomposition efficiency is low in the decomposition process. A conditioning process is necessary to concentrate at a high concentration. For effective treatment for Organic Decontamination Liquid Waste, the composition of organic liquid waste and conditioning process were analyzed. Organic acids, metal ions, radioactive nuclides, surfactants, etc. are present in the Organic Decontamination Liquid Waste, and suspended solids are sometimes generated by various reactions. According to previous studies, the concentration of organic acids including surfactants obtained results from several tens of ppm to a maximum of 1,000 ppm, so the maximum value of 1,000 ppm was assumed. For the composition and total amount of metal ions, the average value (52.7wt% Fe, 16.3wt% Ni, 15.1wt% Cr, 15.9wt% Mn) of the distribution of metal species removed by the actual decontamination process is applied, and the total amount is 1,000 ppm was assumed. As for the radionuclides, only 60Co and 137Cs, which are expected to be mainly present, were considered, and 60Co was assumed to be 2,000 Bq/g and 137Cs to be 360 Bq/g by referring to the literature. The amounts of suspended solids were assumed to be 500 ppm by referring to the characteristics of the liquid waste generated in the decontamination process of the NPPs. Based on the estimated value, a reaction formula was established and a simulated Organic Decontamination Liquid Waste was prepared. As a result of measurement using an analysis device, the composition of the estimated and simulated Organic Decontamination Liquid Waste had similar values. The conditioning and treatment process largely consists of pretreatment, conditioning, decomposition processes. Organic Decontamination Liquid Waste goes through a pretreatment process to remove impurities with large particles. In the conditioning process, treated water that has passed through the UF/RO membrane system is discharged into the environment. At this time, Concentrated water goes through a decomposition process for processing the Organic Decontamination Liquid Waste, and is discharged to the environment through a secondary RO membrane system. The conditioning process is the low-concentration Organic Decontamination Liquid Waste in the UF membrane system is forming a micelles in an RO membrane system, concentrating it to a high concentration and then go through a recirculation process in the UF membrane system. An experiment was conducted to confirm whether the concentration of surfactants occurred during the conditioning process. As a result of the experiment confirmed that the highly concentrated surfactant formed micelles and was filtered out in the UF membrane system.

In Korea, many characteristic component facilities and technologies in general experimental areas for non-radiative materials are owned by industry-academia research. Still, no characteristic analysis test technology has been developed for large, intermediate-level decommissioning waste emitted by neutron irradiation. Since Korea plans to decommission nuclear power plants in 2027, securing analysis technology for intermediate-level decommissioning waste is essential. Accordingly, the Korea Research Institute of Decommissioning (KRID) plans to secure an infrastructure (hot cell) to analyze the characteristics of intermediate-level dismantled waste. Afterward, we intend to stably dispose of the waste generated while decommissioning the current Gori Unit 1/Wolseong Unit 1 using the intermediatelevel dedicated hot cell. It aims to secure high-dose/high-radiation decommissioning waste handling technology through intermediate-level hot cells for the first time in Korea, supports domestic nucleardecommissioning projects, and secure and validate procedures related to material characteristics and nuclide analysis of intermediate-level waste. Furthermore, research on intermediate-level radioactive materials is expected to be carried out in cooperation with schools and research institutes.

이 글은 남한 동해안에 유입된 북한 생활쓰레기로부터 시작된다. 그동안 북한 쓰레기가 남한에 유입된다는 사실은 해양쓰레기 실태조사 관련 연구 에서 주로 이루어졌다. 환동해권 해양쓰레기 유입 등에 관한 연구에서도 주로 한국과 일본, 극동러시아 등의 유입 현황에 관한 연구는 있지만, 북 한은 논의의 대상에서 제외되었다. 본 연구는 기존의 해양학 관점이 아닌 북한학 관점에서 남한 해안에 유입된 북한 쓰레기 문제를 다루었다. 북한 생활쓰레기 중 상품포장지는 직접적으로 북한 상품 생산 현황과 브랜드 등을 알 수 있으며, 간접적으로는 북한 내 경제 상황과 상품 유통 지역망 등을 알 수 있기 때문이다. 동해안 지역에서 수거한 북한제품 포장지를 살 펴보면 대부분 생산공장은 평양으로 표기되었다. 이를 통해 평양에서 생산 한 제품이 동해안 지역으로 유통됨을 알 수 있다. 또한 동해안 지역은 북 한을 대표하는 대도시인 원산, 청진, 함흥, 라선 등이 있는데, 실제로 상품 포장지에는 이 지역 생산공장이 표기되는 사례도 있었다. 김정은 집권 이 후 매년 국가적 차원에서 국가디자인전시회를 개최할 만큼 산업미술을 강 조하는데, 특히 상품의 고유한 특징을 표현하는 상표도안을 강조한다. 본 연구에서는 동해안 주요 도시에서 생산한 제품을 중심으로 같은 품목이지 만 공장별로 어떻게 상표도안이 다른지 살펴봤다. 북한 쓰레기에 대한 북 한학적 시각과 해양학적 시각의 학제간 연구를 통해 남한에 유입되는 북 한 생활쓰레기에 대한 연구의 폭을 넓혀갈 필요가 있다.

PURPOSES : The purpose of this study is to confirm the thermal expansion characteristics of concrete mixed with 1% waste glass fine aggregates, which is the amount stipulated for recycled aggregates in the current quality standard. METHODS : The coefficient of thermal expansion was measured by applying AASHTOT 336-10 using a LVDT. The results measured were used as physical properties in a finite element analysis to confirm the change in tensile stress and the displacement of the right angle section of the upper slab of a concrete pavement due to admixture substitution. RESULTS : The thermal expansion coefficients of concrete based on the replacement rate of the admixture when the waste glass fine aggregates are replaced are within the range of the thermal expansion coefficients of concrete specified in the Federal Highway Administration report. As the replacement rate of the admixture increases, the thermal expansion coefficient of concrete decreases. As the thermal expansion coefficient decreases, the slab pavement curling displacement and the tensile stress of the center of the upper slab of concrete decrease. CONCLUSIONS : In the short term, the presence or absence of waste glass fine aggregates does not significantly affect the thermal expansion coefficient of concrete. However, in the long term, waste glass fine aggregates are reactive aggregates that causes ASR, which creates an expandable gel around the aggregates and results in concrete expansion. Therefore, the relationship between ASR and the thermal expansion coefficient must be analyzed in future studies.

In a nuclear power plant, the activated corrosion products are deposited on the reactor coolant system. The activated corrosion products must be removed to reduce the radiation exposure to workers before maintaining or decommissioning of the nuclear power plant. In order to remove the remove the activated duplex oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR), the Cyclic SP (Sulfuric acid/Permanganate)-HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process developed by KAERI (Korea Atomic Energy Research Institute) can be used. After applying the Cyclic SP-HyBRID process, a sulfate-rich waste powder containing the radionuclide is generated, and the radioactive powder has to be stabilized for final disposal. In the previous study, it was confirmed that the low-temperature sintering method can be applied to immobilize the sulfate-rich waste powder. Thus, immobilization of the Cyclic SP-HyBRID process waste powder was carried out by the low-temperature sintering method using a low melting point glass, and the physicochemical and radiological characteristics of a waste form were evaluated in this study. As a result, the compressive strength of the waste form increased with increasing sintering temperature and sintering time. It is considered that the result was caused by the difference in the band gap between the bismuth borate and zinc borate, which are the products during the sintering process. It was verified that the physical stability was maintained after the 107 Gy of irradiation test. In addition, it was confirmed that the radioactive metal hydroxides contained in the waste powder converted to metal oxide forms, which have the lower solubility, at the sintering temperature. Finally, the waste form was evaluated as a low-level radioactive waste from the concentration of radionuclides contained in the waste form.

In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

The structural integrity of concrete silos is important from the perspective of long-term operation of radioactive waste repository. Recently, the application of acoustic emission (AE) is considered as a promising technology for the systematic real-time health monitoring of concrete-like brittle material. In this study, the characteristics of AE wave propagation through concrete silo of Gyeongju radioactive waste repository were evaluated under the effects of groundwater and temperature for the quantitative damage assessment. The attenuation coefficients and absolute energies of AE waves were measured for the temperature cases of 15, 45, 75°C under dry and saturated concrete specimens, which were manufactured based on the concrete mix same as that of Gyeongju concrete silo. The geometric spreading and material loss were taken into account with regard to the wave attenuation coefficient. The attenuation coefficient shows a decreasing pattern with temperature rise for both dry and saturated specimens. The AE waves in saturated condition attenuate faster than those in dry condition. It is found that the effect of water content has a greater impact on the wave attenuation than the temperature. The results from this study will be used as valuable information for estimating the quantitative damage at the location micro-cracks are generated rather than the AE sensor location.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

최근 탄소 중립 정책에 따른 신재생에너지 활용을 위한 노력이 가속화되고 있다. 이를 위하여 본 연구에서는 바이오매스 작물 중 하나인 케나프 (Hibiscus cannabinus L.)를 연료화하기 위하여, 미이용 목재 자원과 폐목재 자원을 혼합하여 고형연료인 펠릿을 제조하고 품질을 분석하였다. 품질을 평가하기 위해 목재 펠릿, 비목재 펠릿과 Bio-SRF의 품질기준을 통해 성형한 펠릿의 품질을 파악하였다. 케나프 펠릿의 경우 겉보기밀도, 발열량 등에서는 목재 펠릿 품질기준을 만족하였으나 회분의 함량이 기준을 초과하였다. 이를 보완하기 위해 목재 자원인 폐목재를 혼합하여 제조한 펠릿의 특성을 조사한 결과, 질소 및 겉보기밀도, 회분, 발열량 등에서 오히려 품질을 저하시키는 경향을 보이는 것으로 나타났다. 한편, 미이용 목재를 혼합하여 성형된 펠릿의 품질을 조사한 결과, 겉보기밀도, 함수율, 질소, 황, 발열량의 조건에서 대부분 목재 펠릿의 품질기준을 만족하였다. 다만 회분함량의 경우 비목재 펠릿 및 Bio-SRF의 15% 이하 기준을 만족하고 있지만, 목재 펠릿의 최저 기준인 B등급 2.0% 이하 기준의 경우 만족하는 경우와 만족하지 못하는 경우가 발생하였다. 함수율 15%(w.b.)에서 케나프와 미이용 목재의 혼합비가 2:8인 경우와 함수율 20%(w.b.)에서 케나프와 미이용 목재의 혼합비가 6:4 및 2:8인 경우에 기준을 만족하였고, 그 이외에는 기준을 만족하지 못하였다. 특히, 케나프만을 사용하거나 폐목을 섞은 경우는 모두 기준을 만족하지 못하므로, 목재 펠릿의 기준을 만족하는 연료 이용을 위해서는 케나프와 미이용 목재 자원을 혼합 활용하는 것이 바람직할 것으로 판단된다.