## 1 Introduction  
Recently, a combined pyrochemical and hydrometallurgical technology (\"PH-process\") has been proposed for processing of the mixed uranium-plutonium nitride spent nuclear fuel (MNUP SNF) of the BREST-OD-300 reactor with lead coolant [1]. Pyrochemical processing includes anodic dissolution of the MNUP SNF in a melt of alkali metal chlorides and subsequent cathodic deposition of fissile materials on a liquid cadmium cathode. During this process, a new type of radioactive waste (RW)-spent electrolyte-is produced. This material, as any other types of RW, must be conditioned to obtain a stable (chemically, mechanically, etc.) compound. The choice of a matrix material for the immobilization of such salt waste of a complex chemical composition is not yet obvious, because the activity of the spent electrolyte--of high or medium level--depends on the decontamination factor from the radionuclides of the processed MNUP SNF, and chlorides are difficult to immobilize compounds. Currently, glass is the only industrial matrix for solidification and immobilization of high-level waste (HLW) used in Russia, Great Britain, Germany, France, USA, China and other countries. Vitrified waste forms have undoubted advantages, first of all, the possibility for continuous control of the technological process, the ability to include various radionuclides, and a relatively high stability. On the other hand, the vitrification is a high temperature process (for phosphate glass (900-1050) \({}^{\circ}\)C [2], for borosilicate glass \(\sim\)1150 \({}^{\circ}\)C) [3]), which results in the loss of volatile radionuclides. Moreover, chlorides have the low solubility in glass materials (for example, in borosilicate glass less than 1.5 wt% [4]). Thus, the search for alternative matrix materials that would reliably include and immobilize radioactive chloride waste and other volatile components is an acute task. The most studied for immobilizing salt waste matrices are the \"zeolite/sodalite in glass\" systems which are used, for example, in the USA for conditioning of the spent electrolyte from the reprocessing of spent nuclear fuel of the EBR-II reactor [5; 6; 7; 8; 9]. Moreover, an alternative method of indirect immobilization was developed by occlusion of a pre-dechlorinated salt waste using a SAP (SiO\({}_{2}\)-Al\({}_{2}\)O\({}_{3}\)-P\({}_{2}\)O\({}_{5}\)) composite with subsequent solidification in a binder glass [10; 11]. Earlier [12; 13], a mineral-like titanate matrix, Synrock, was proposed. The main minerals of Synrock are zirconolite, hollandite, perovskite, and rutile. Synroc can be synthesized by hot isostatic pressing and sintering methods. The last was used to produce ceramics for immobilization the Cs and Sr fractions of HLW. The possibility of converting alkali metal chlorides into orthophosphate ceramics was studied by using kosnarite, NZP (NaZr\({}_{2}\)(PO\({}_{4}\))\({}_{3}\)), and langbeinite as matrix materials with addition of NaF as a binder to increase chemical resistance [14]. There is a technology to manufacture fused ceramics that is similar to the production of glassy materials, and a number of works consider murataite as a promising material [15; 16]. In the pyrochemical processing of SNF, the murataite matrix is proposed to be used for immobilization of HLW generated during anhydrous reprocessing of SNF of fast neutron reactors, as well as during calcination of solutions used for decontamination of glove boxes and hot cells [16]. In both cases, metal oxides (TiO\({}_{2}\), MnO\({}_{2}\), CaO, Al\({}_{2}\)O\({}_{3}\), Fe\({}_{2}\)O\({}_{3}\), ZrO\({}_{2}\)) were used as the main components to synthesize murataite. Sodium aluminophosphate (NAFP) glass is a promising matrix for the immobilization of HLW, since its synthesis temperature is lower than that of aluminophosphate glass, which is the only industrial matrix for HLW immobilization used in Russia [2].