In the paper, a novel technique for highly dispersed pyrochlore Y₂Ti₂O₇ is proposed. The experimental results proved that the application of microwave irradiation at a certain stage of calcination allowed synthesizing of Y₂Ti₂O₇ in much shorter time, which ensured substantial energy savings. An increase up to 98 wt.% in the content of the preferred phase with a pyrochlore-type structure Y₂Ti₂O₇ was obtained after 25 h of yttrium and titanium oxides calcination at a relatively low temperature of 1150 °C, while the microwave-supported process took only 9 h and provided 99 wt.% of pyrochlore. The proposed technology is suitable for industrial applications, enabling the fabrication of large industrial amounts of pyrochlore without solvent chemistry and high-energy mills. It reduced the cost of both equipment and energy and made the process more environmentally friendly. The particle size and morphology did not change significantly; therefore, the microwave-assisted method can fully replace the traditional one.

The results of experimental research for obtaining of phosphate matrix materials: fluorapatite Ca10(PO4)6F2, sodium zirconium phosphate NaZr2(PO4)3 and potassium magnesium phosphate KMgPO4·6H2O were presented. The evolution of phase composition in their synthesis was investigated. The optimal parameters for obtaining monophasic phosphate matrix materials were found. Possibility of obtaining powders of calcium Ca10(PO4)6F2 and strontium-containing Ca9Sr(PO4)6F2 fluorapatites by both solid phase reaction method with subsequent heat treatment and chemical co-precipitation method from solutions of the initial components was investigated. Nanosized zirconium orthophosphate powders NaZr2(PO4)3 were synthesized by the sol-gel method. Hydrated phosphatic matrices KMgPO4·6H2O by a chemical reaction between MgO and KH2PO4 in water at a room temperature were obtained. The requirements for high-level waste matrix were presented. The suitability of the use of synthetic phosphate materials as a matrix for the immobilization of high level nuclear wastes was determined.

Magnesium potassium phosphate cements (MKPCs) are currently considered as the matrix materials for immobilization of radioactive waste of low and medium activity, as an alternative to Portland cements. At the same time, MKPC compound has the following advantages: fast hardening, high early strength, low shrinkage and high chemical resistance. Magnesium potassium phosphate cement is prepared at room temperature by an acid-base reaction between calcined magnesium oxide (MgO), potassium dihydrogen phosphate (KH2PO4) and water to form magnesium potassium phosphate hexahydrate MgKPO4 ∙ 6H2O known as K-struvite. To obtain MKPC, starting components of high purity and, accordingly, high cost are required. From an industrial point of view, this leads to an excessive price of the final product, and therefore, the use of inexpensive industrial waste as cement fillers is economically cost effective. In the present paper, the effect of fly ash and blast-furnace slag additives on the microstructure, compressive strength, and chemical resistance of MKPC samples was studied. The results indicate that particles of both ash and blast-furnace slag are involved in the reaction of MKPC obtaining with the formation of predominantly crystalline K-struvite. According to X-ray diffraction analysis, the content of K-struvite in samples with additives of both fly ash and blast-furnace slag is almost the same and amounts to 58–60 %. With the same addition of filler, samples with the addition of blast-furnace slag have a denser structure, which is formed due to the increased reactivity of blast-furnace slag particles. Moreover, MKPC samples with the addition of blast-furnace slag exhibit higher strength and better chemical resistance to leaching compared to MKPC samples with fly ash additions. The obtained results demonstrate the prospects of using industrial waste additives in the production of MKPC compounds characterized by high mechanical strength and chemical resistance. Магній-калій-фосфатні цементи (МКФЦ) розглядаються зараз як матеріали матриць для іммобілізації радіоактивних відходів низької та середньої активності, як альтернатива портландцементу. Водночас MKФЦ компаунд має такі переваги: швидке твердіння, висока рання міцність, малі усадки та висока хімічна стійкість. МКФЦ одержують за кімнатної температури внаслідок кислотно-лужної реакції між обпаленим оксидом магнію (MgO), дигідрофосфатом калію (KH2PO4) та водою з утворенням гексагідрату фосфату магнію та калію MgKPO4·6H2O, відомого як К-струвіт. Для отримання МКФЦ потрібні вихідні компоненти високої чистоти та, відповідно, високої вартості. З промислового погляду, це призводить до надмірної ціни кінцевого продукту і тому використання недорогих промислових відходів, як наповнювачів цементів, є економічно виправданим. У цій статті досліджено вплив добавок золи виносу та доменного шлаку на мікроструктуру, міцність на стиск та хімічну стійкість МКФЦ зразків. Наведені дані свідчать про те, що в реакції отримання МКФЦ беруть участь як частинки золи, так і доменного шлаку з формуванням переважно кристалічного К-струвіту. За даними рентгеноструктурного аналізу вміст К-струвіту у зразках із добавками, як золи виносу, так і доменного шлаку практично однаковий і становить 58 – 60 %. У разі однакового додавання наповнювача зразки з добавкою доменного шлаку мають більш щільнішу структуру, яка утворюється завдяки підвищеній реакційній здатності частинок доменного шлаку. Крім того, МКФЦ зразки з додаванням доменного шлаку демонструють більш високу міцність та кращу хімічну стійкість до вилуговування порівняно з МКФЦ зразками з добавками золи виносу. Отримані результати свідчать про перспективність використання добавок промислових відходів у виробництві МКФЦ компаундів, що характеризуються високими значеннями механічної міцності та хімічної стійкості.

The study on the posibility of the use of the optimal glass compositions for vitrification of an imitator of vat residues of liquid radioactive waste from nuclear power plants with VVER-1000 reactors was carried out. The main process parameters such as vitrification temperature, strength, corrosion resistance, absence of crystalline phases, minimization of glass-forming additives and inclusion the maximum amount of waste were analyzed. It has been established that the melting temperature of lead-borosilicate glass matrices was 1150 °C, which satisfies the requirements for vitrification of low- and medium-level waste. The ultimate compressive strength of the obtained samples of glass matrices was 136.0 MPa. In addition, it has been shown that lead-borosilicate glass matrices are the most resistant to leaching. The cesium leaching rate was 1.5·10-5 g/cm2·day.

Several recent studies have evaluated technologies of spent nuclear fuel processing specifically for solidifying transuranic (TRU) waste as a by-product of fission. Of the TRU group, plutonium and the minor actinides will be responsible for the bulk of the radiotoxicity and heat generation of spent nuclear fuel in the long term (300 to 20,000 years). In this study, we investigated magnesium potassium phosphate (MKP)-based compounds as host waste forms for the encapsulation of inactive trivalent Nd and Sm as analogues of the minor trivalent actinides, Am and Cm. Waste forms were fabricated under ambient atmospheric conditions by adding 5 wt.% of substances containing Nd or Sm via the following two routes: powder oxides and aqueous solutions of nitrate salts. Waste form performance was established using strength and aqueous medium leaching tests of MKP-based specimens. The MKP materials were analyzed by X-ray diffraction (XRD), scanning and transmission electron microscopy (SEM and TEM), energy-dispersive X-ray spectroscopy (EDS), and Raman spectroscopy. The waste forms exhibited a compressive strength of ≥30 MPa and were durable in an aqueous environment. The leachability indices for Nd and Sm, as per the ANS 16.1 procedure, were 19.55–19.78 and 19.74–19.89, respectively, which satisfy the acceptable criteria (>6). The results of the present room temperature leaching study suggest that MKPs can be effectively used as a host material to immobilize actinides (Am and Cm) contained in TRU waste.

The state of liquid radioactive wastes (LRW) management at Ukrainian Nuclear Power Plants (NPPs) is characterized by the lack of a completed technological cycle from processing to obtaining the final product suitable for further long-term storage or disposal. As a result, the storage tanks for bottoms residue (BR) are 65-75% full (Zaporozhye and South-Ukrainian NPPs), and the resource for placing molten salt at Zaporizhzhya NPP (92.7%) is close to exhaustion [1]. Therefore, the development of technologies and materials for NPP LRW solidification is an urgent need and aims to ensure the processing of LRW to a solid state that will meet the acceptance criteria for disposal in centralized storage facilities. One of the effective methods of LRW solidification is their vitrification. The main advantage of vitrification is that during the vitrification process the volume of waste is reduced by several times and this saves expensive storage space [2, 3]. The purpose of this work is to calculate the absorbed dose that borosilicate glass matrices with included bottoms residue will accumulate over 300 years of storage, and to study the effect of simulated X-ray irradiation on their physical and mechanical properties.