Vitrification, one of the most promising solidification processes for various materials, has been applied to radioactive waste to improve its disposal stability and reduce its volume. Because the thermal decomposition of dry active waste (DAW) significantly reduces its volume, the volume reduction factor of DAW vitrification is high. The KHNP developed the optimal glass composition for the vitrification of DAW. Since vitrification offers a high-volume reduction ratio, it is expected that disposal costs could be greatly reduced by the use of such technology. The DG-2 glass composition was developed to vitrify DAW. During the maintenance of nuclear power plants, metals containing paper, clothes, and wood are generated. ZrO2 and HfO2 are generally considered to be network-formers in borosilicate-based glasses. In this study, a feasibility study of vitrification for DAW that contains metal particulates is conducted to understand the applicability of this process under various conditions. The physicochemical properties are characterized to assess the applicability of candidate glass compositions.
The immobilization of low- and intermediate-level radioactive waste (LILW) is crucial for its final disposal in repositories. While cementitious waste forms have conventionally been used for immobilizing various LILWs, they suffer from several issues, including poor durability, low resistance to leaching, and limited waste loading capacity. As an alternative, alkali or acid-activated geopolymer waste forms have garnered global attention. Unlike cementitious waste forms, geopolymer waste forms exhibit excellent physicochemical characteristics due to their three-dimensional amorphous structure and low calcium content. In this work, we provide an overview of geopolymer waste form research being conducted in countries such as Japan, the United Kingdom, the European Union, and South Korea. We specifically focus on the immobilization of soil waste, spent ion exchange resins, organic liquid waste, and evaporator concentrate (borate waste). We also identify the factors influencing the physicochemical characteristics of geopolymer waste forms and their immobilization performance. We propose a guide for optimizing the molar mixing formulations of geopolymer waste forms, including the selection of appropriate precursor materials. Additionally, we discuss the future prospects and significant challenges in the field of geopolymer waste forms that need to be addressed for their application in radioactive waste management.
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.
A solution combustion process for the synthesis of hollandite (BaAl2Ti6O16) powders is described. SYNROC (synthetic rock) consists of four main titanate phases: perovskite, zirconolite, hollandite and rutile. Hollandite is one of the crystalline host matrices used for the disposal of high-level radioactive wastes because it immobilizes Sr and Lns elements by forming solid solutions. The solution combustion synthesis, which is a self-sustaining oxi-reduction reaction between a nitrate and organic fuel, generates an exothermic reaction and that heat converts the precursors into their corresponding oxide products in air. The process has high energy efficiency, fast heating rates, short reaction times, and high compositional homogeneity. To confirm the combustion synthesis reaction, FT-IR analysis was conducted using glycine with a carboxyl group and an amine as fuel to observe its bonding with metal element in the nitrate. TG-DTA, X-ray diffraction analysis, SEM and EDS were performed to confirm the formed phases and morphology. Powders with an uncontrolled shape were obtained through a general oxide-route process, confirming hollandite powders with micro-sized soft agglomerates consisting of nano-sized primary particles can be prepared using these methods.
The soils contaminated with radionuclides such as Cs-137 and Sr-90 should be solidified using a binder matrix, because radioactively contaminated soils pose environmental concerns and human health problems. Ordinary Portland cement has been widely used to solidify various radioactive wastes due to its low cost and simple process. In this study, simulant soil waste was solidified using cement waste form. The soils were collected around ‘Kori Nuclear Power Plant Unit 1’ and they were contaminated with the prepared simulant liquid waste containing Fe, Cr, Cs, Ni, Co, and Mn. The water-to-dry ingredients (W/D) ratio of cement waste form was 0.40. The cement paste was poured into a cubic mold (5×5×5 cm) and then cured for 28 days at room temperature. The 28-day compressive strength, water immersion, and EPA1311-toxicity characteristic leaching procedure (TCLP) tests were performed to evaluate the structural stability of cement waste form. The compressive strength was not proportional to soil waste loading, and the lowest compressive strength (4±0.1 MPa) was achieved in cement waste form containing 50wt% soil waste. After the water immersion test for 90 days, the compressive strength of cement waste form with 50wt% soil waste increased to 7.5±0.6 MPa, meeting the waste form acceptance criteria in the repository. It is believed that long-term water immersion test contributed to the additional curing and hydration reaction, resulting in the enhanced compressive strength. As a result of the TCLP test, the released amount of As, Ba, Cd, Cr, Pb, Se, Co, Cs, and Sr was less than the domestic and international standards. These results imply that cement waste form can be a promising candidate for the solidification of radioactive soil wastes.
Immobilization of radioactive borate waste containing a high boron concentration using cement waste form has been challenged because the soluble borate phase such as boric acid reacts with calcium compounds, hindering the hydration reaction in cement waste form. Metakaolin-based geopolymer waste form which has a pure aluminosilicate system without calcium can be a promising alternative for the cement; however, secondary B-O-Si networks are formed by a reaction between borate and silicate, resulting in poor mechanical characteristics such as low compressive strength and final setting retardation. Thus, it is important to optimize the Si/Al molar ratio and curing temperature which are critical parameters of geopolymer waste form to increase borate waste loading and enhance the durability of geopolymer. Here, metakaolin-based geopolymer waste form to immobilize simulant radioactive borate waste was fabricated by varying the Si/Al molar ratio and curing temperature. The 7 days-compressive strength results reveals that the Si/Al molar ratio of 1.4 and curing at 60°C is advantageous to achieving high waste loading (30wt%). In addition, geopolymer waste forms with the highest borate waste loading exceeded the 3.445 MPa after the waste form acceptance criteria such as thermal cycling, gamma irradiation, and water immersion tests. The leachability index of boron was 7.56 and the controlling leaching mechanism was diffusion. The thermal cycling and gamma irradiation did not significantly change the geopolymer structure. The physically incorporated borate waste was leached out from geopolymer waste form during leaching and water immersion tests. Considering these results, metakaolin-based geopolymer waste form with a low Si/Al ratio is a promising candidate for borate waste immobilization, which has been difficult using cement.
Two waste forms, namely cement and geopolymer, were investigated and tested in this study to solidify the corrosive sludge generated from the surface and precipitates of the tubes of steam generators in nuclear power plants. The compressive strength of the cement waste form cured for 28 days was inversely proportional to waste loading (24.4 MPa for 0wt% to 2.7 MPa for 60wt%). The corrosive sludge absorbed the free water in the hydration reaction to decrease the cementation reaction. When the corrosive sludge waste loading increased to 60wt%, the cement waste form showed decreased compressive strength (2.7 MPa), which did not satisfy the acceptance criteria of the repository (3.45 MPa). Meanwhile, the compressive strength of the geopolymer waste form cured for 7 days was proportional to waste loading (23.6 MPa for 0wt% to 31.9 MPa for 40wt%). The corrosive sludge absorbed the free water in the geopolymer when the water content decreased, such that a compact geopolymer structure could be obtained. Consequently, the geopolymer waste forms generally showed higher compressive strengths than cement waste forms.
Three kinds of STS304-Zr alloys were fabricated by varying the Zr content, and their microstructure and fracture properties were analyzed. Moreover, we performed heat treatment to improve their properties and studied their microstructure and fracture properties. The microstructure of the STS304-Zr alloys before and after the heat treatment process consisted of α-Fe and intermetallics: Zr(Cr, Ni, Fe)2 and Zr6Fe23. The volume fraction of the intermetallics increased with an increasing Zr content. The 11Zr specimen exhibited the lowest hardness and fine dimples and cleavage facets in a fractured surface. The 15Zr specimen had high hardness and fine cleavage facets. The 19Zr specimen had the highest hardness and large cleavage facets. After the heat treatment process, the intermetallics were spheroidized and their volume fraction increased. In addition, the specimens after the heat treatment process, the Laves phase (Zr(Cr, Ni, Fe) 2) decreased, the Zr6Fe23 phase increased and the Ni concentration in the intermetallics decreased. The hardness of all the specimens after the heat treatment process decreased because of the dislocations and residual stresses in α-Fe, and the fine lamellar shaped eutectic microstructures changed into large α-Fe and spheroidized intermetallics. The cleavage facet size increased because of the decomposition of the fine lamellarshaped eutectic microstructures and the increase in spheroidized intermetallics.
A full-scale process has been developed to immobilize fission products that accumulate within the Mark IV electrorefiner (ER) electrolyte at Idaho National Laboratory. ER salt was blended with treatment additives, followed by pressureless consolidation (PC) in a furnace to produce a durable ceramic waste form (CWF). The goal is the development of a process to consolidate actual radioactive ER salt into a form suitable for transportation and disposal.Four batches (300 to 400 kg per batch) of full-scale pre-qualification material preparation runs have been prepared. From these four batches of nonradioactive salt-loaded surrogate material, three full-scale PC trials have been conducted. The first PC test run, established equipment parameters with a basic CWF container design. The second trial included a modified CWF container design, real-time measurement of CWF consolidation, and an audio recording to identify cracking during the CWF cool-down. During the third trial, salt was doped (from the fourth material preparation batch) to create a nonradioactive salt material and to more closely represent actual ER salt. The second and third trials were also used to validate a model developed for the CWF. The CWF model is beneficial for understanding and predicting the physical processes that occur during the heat cycle. This would be particularly useful when the CWF is located in a hot cell, which makes accessing and examining a CWF difficult.
붕산폐액을 함유한 시멘트 및 파라핀 고화체, 폐이온교환수지를 함유한 시멘트 고화체 그리고 잡고체중의 제염지에 대하여 Co-60을 조사선원으로 하여 rads까지 조사하여 발생되는 분해가스의 종류 및 그의 발생량을 분석하였다. 그 결과 분해가스로는 및 등이 발생하였으며, 가 대부분을 차지하였다. 가스발생량은 폐기물과 고화매질의 종류에 따라 으로 상당한 차이를 보였으며, 폐이온교환수지를 함유한 고화체에서 가장 높은 분해가스 발생량을 보였다. 그리고 수소가스는 제염지 폐기물에서 가장 많이 발생하였다. 제염지의 는 0.12이었다.
저방사성폐수지, 제올라이트, 가연성잡고체 혼합폐기물을 유리화하기 위해서 AG8W1 후보 유리와 가연성잡고체 단독으로 유리화하기 위하여 DG-2 후보유리가 개발되었다. 두 후보유리의 화학적 내구성을 평가하기 위하여 PCT와 VHT 침출시험이 수행되었다. 7일 PCT 침출시험 수행결과 AG8W1과 DG-2의 주요 원소별 침출률은 기준유리(benchmark glass SRL-EA) 보다 낮게 나타남을 알 수 있었고 미국 Hanford 유리고화체 규제치 보다 낮은 결과를 나타냄을 알 수 있었다. 또한, 120일 동안의 시험에서도 주요 원소인 B, Na, Si, Li가 SRL-EA 보다 낮게 나타남을 볼 수 있었다. VHT 침출시험 수행결과 AG8W1, DG-2의 침출률(leach rate)은 각각 , 로써 미국 Hanford 규제치 보다 낮은 결과를 나타냈다. 결과적으로 유리화 시설 상용운전 시 사용예정인 이들 후보유리들의 침출은 안정화되어 있었으며 화학적 내구성이 우수함을 알 수 있었다.
원자력발전소 해체과정에서 발생하는 폐기물 중 가장 큰 비중을 차지하는 것은 콘크리트 재료이다. 일반적적으로 콘크리트의 방사화는 시멘트 페이스트에 집중되기 때문에 방사화된 시멘트 페이스트만 고화처리를 하고 굵은 골재를 재활용할 수 있다면 원자력발전소 해체 과정에서 발생하는 폐기물의 양을 현저하게 감소시킬 수 있다. 본 연구에서는 저준위 방사화 콘크리트의 처리방안 개발을 위한 기초연구를 목적으로 수행되었으며, 이를 위하여 20년 이상 경과한 폐콘크리트에서 분리한 미분말을 고화처리하는 방법에 대하여 검토하고자 하였다.
최근 환경적 · 사회적으로 문제가 되고 있는 산업폐기물을 지반공학적 재료로 재활용하기 위한 관심이 확대대고 있는 추세이다. 따라서 본 연구에서는 화력발전소의 대표적인 산업부산물인 석탄회 중 저회의 도로 성토용 재료 및 구조물 뒤채움용 재료로의 이용을 위해 폐어망보강 저회의 CBR 특성을 분석하였다. 폐어망의 보강 방법은 지오그리드와 같은 층보강 형태, 그리고 단섬유처럼 불특정보강 형태를 이용하였고, 지지력 시험 결과 CBR 값은 짧게 잘라서 랜덤하게 혼합한 경우보다 층으로 보강하였을 때 더 높은 결과 값을 나타내었다. 또한 보강 층수가 증가할수록 보강효과도 증가하는 경향을 보였다.