Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination of possible chemical compounds in which 14C can be found and assessment of fastness of its binding in the structure of irradiated graphite. Indigent and foreign experience of handling graphite radioactive wastes was analyzed, calculations and measurements were performed. Information was provided on the channels of accumulation of 14C in the structure of reactor graphite and it was demonstrated that the largest quantities of the radionuclide in question are generated according to the reaction 14N(n, p)14C. Here, most part of radioactive carbon is generated on 14N nuclei found in the form of impurities in non-irradiated graphite and in the composition of gas used for purging nuclear reactor in the process of operation. 14C radionuclide generated according to 14N(n, p)14C nuclear reaction is localized in the near subsurface graphite layer (in the near subsurface layer of pores) at the depth of not more than 50 nm. Analysis was performed of possible chemical compounds which may incorporate radioactive carbon. It was established that the form of occurrence is determined by the operational properties of specific graphite element in the reactor core. 14C binding energy in the structure of irradiated graphite was evaluated and depth of its penetration in the structure was calculated. It was established that selective extraction of this radionuclide is possible only under elevated temperatures in weakly oxidizing environment which is explained by the binding energy reaching up to 800 kJ/mole in the process of chemical sorption of 14C on the surface of graphite and depth of its occurrence equal to ~ 70 nm in the course of ion implantation. It was demonstrated that radioactive carbon generated according to 13C(n, γ)14C nuclear reaction is uniformly distributed among graphite elements and possesses binding energy ~477 kJ/mole. Its selective extraction is possible only under the condition of destruction of graphite crystalline lattice and organization of the process of isotopic separation. The obtained results allow recommending the most efficient methods of handling irradiated graphite during decommissioning uranium-graphite reactors.
A factorial study is made of precipitation of ammonium polyuranates from nitric acid solutions of uranium in the presence of urea, including measuring the effects of the temperature and urea and uranium concentrations in the initial solution on the sedimentation and filtration characteristics of the precipitates. The conditions of the process are optimized to obtain crystalline, readily filterable precipitates.Precipitation of ammonium polyuranates with ammonia from concentrated uranium backwash solutions finds wide use in the commonly accepted extraction3 precipitation reprocessing of spent nuclear fuel [1,2]. That is why intensive studies have been made of the mechanism of precipitation of ammonium polyuranates and of the properties of the resulting precipitates as influenced by the precipitation conditions [3,4].In the extraction3precipitation reprocessing of highly enriched (weapon-grade) uranium at the Siberian Chemical Combine, ammonium polyuranates are precipitated from solutions containing considerable concentrations of urea (urea is used in the backwashing stage, to obtain uranium concentrates) [5,6].There are only limited data in the literature on precipitation of ammonium polyuranates in the presence of urea [7]. Therefore, in this work we thoroughly studied the effect of the urea concentration in nitric acid solutions of uranium on the precipitation of ammonium polyuranates with ammonia. EXPERIMENTAL Stock uranium solutions were prepared by dissolving uranyl nitrate hexahydrate in 0.2 M HNO 3 [ultrapure grade, OST (Branch Standard) K-03-265376]. Then, a fixed amount of urea [GOST (State Standard) 2081392, grade A] was dissolved in the uranium solution.Precipitation of ammonium polyuranates from urea-containing uranyl nitrate solutions was carried out in a continuous precipitator with automatic control of the temperature and pH. Ammonia (25 wt % solution, GOST 6221-82E) and the stock uranium solution were fed simultaneously to the precipitator (slurry residence time in the reactor 1 h). The flow rate of the initial solution was controlled using a batcher, and that of the ammonia solution was automatically controlled by pH. Fixed pH was maintained with a BAT-15 automatic titration unit connected to a pH-121 pH meter (ESL 63-07 pH-metric glass electrode; EVL-1MB Ag/AgCl reference electrode). Stirring was carried out with a propeller mixer. The temperature control was realized using a contact thermometer (GOST 9871361) connected to a temperature controller. Slurry was transferred through an upper discharge to a settling tank. After 3 h of continuous operation of the system, samples of the slurry were taken, and the aqueous phase composition and physicochemical characteristics of the precipitates were analyzed.The completion of precipitation was characterized by the uranium concentration in the mother liquors, and properties of the precipitate, by its moisture content and permeability coefficient. The relative volume of the precipitate was estimated as the ratio of the precipitate volume after settling to ...
The article presents methods and a general survey procedure for storage facilities with radioactive waste generated during the operation of uranium-graphite reactors. It presents the experience from the application of different survey methods and relevant findings. The paper describes the key areas for further improvement of the survey methods, including the results from the mock-up tests based on various methods applied for arranging penetrations in the radioactive waste volume providing access to the bottom waste layers. The approaches presented in the article can be adapted to explore other storage facility types arranged during nuclear decommissioning.
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