2^nd International Conference on Nuclear Chemistry November 15-16, 2017
(10 Plenary Forums - 1Event)
Las Vegas, Nevada, USA

Andrey A. Naumov is working in the laboratory of spent nuclear fuel reprocessing of the Khlopin Radium Institute (St-Petersnurg, Russia). He is a member of the scientific team which have dealt with investigation, development and testing of modern extraction for aqueous processing of irradiated materials of different kind. His main scientific interest now is to study the extraction of molybdenum-99 from solutions of irradiated uranium targets using higher hydroxamic acids, applicating for PhD in radiochemistry.

Extraction of molybdenum, uranium, and certain fission products from nitric acid solutions with 0.2% solutions of higher hydroxamic acids (HA) in alcohols poorly soluble in water has been studied for 99Mo concentrate production from solutions of enriched U targets, including those of very low enrichment (3% 235U dioxide). \r\nThe process chemistry of Mo has been supposed for both steps of Mo interphase transfer. Mo extraction by the solvent containing HA, namely caprinohydroxamic acid (CHA), is a slow process limited by first-order reaction in aqueous phase. The slope of DMo on its aqueous concentration indicates on apparent coordination number ~ 3 (or more exactly, ~2,7 due to Mo partial dimerization in the solvent phase), D Mo rising with the length of alcohol alkyl radical.. One alcohol molecule might enter into the Mo complex with HA.\r\nMo backwashing has been performed by interphase autocatalytic thermo-chemical HA oxidizing destruction with 5 mol/L HNO3 at 95 oC in combination with Mo complexing in aqueous phase by a salt-free reagents. Alkaline scrubbing has been chosen for final regeneration of the solvent base. \r\nThe final rig trial included (see Fig.) dissolution of U-Al model or real targets in 8 mol/L HNO3 containing 0.2 g/L Hg and 0.2 g/L HF at 95oC, allowing further radioiodine and ruthenium compounds air stripping. Batch extraction process of Mo recovery was carried out after cooling of the feed using 27 mmol/L caprinohydroxamic acid in 20% n-decanol with Isopar-M in 3 steps – extraction, scrubbing and backwashing - in the vessels of decreasing volume according to the concentrating factor. The simulate feed contained, mol/L : HNO3 – 1,2; Al – 1,2; Fe – 5∙10-3; U – 0,11; Hg – 1∙10-3, 239Pu –1,4∙10-4, Mo – 3,2∙10-5, as well as 15 MBq/L 99Mo, 5,2 MBq/L 125I and 4,1 MBq/L 239Np.\r\nThe achieved concentrating factor was 180 in total, including 18 at extraction and 10 at backwashing at 90oC. Total process duration was 2 h. Decontamination factors were ~ 1.5•106 from U, ~ 850 from 125I, > 105 from 239Pu, > 106 from 239Np, > 106 from Al, 4.6∙104 from Fe, ~ 2•104 from Hg. \r\nThe compact extraction flowsheet and simple equipment are proposed. Final Mo decontamination could be performed by sorption and/or by sublimation. The method could be assumed as an interlocutory decision between HEU irradiation and that one of high purity Mo even isotopes (98Mo and 100Mo).

Egor A. Puzikov graduated from Physico-Chemical Department of SPb Institute of Technology (Technical University), St-Petersburg, Russia, in 1989 and received there his PhD in 1995. He has been working at Khlopin Radium Institute for 23 years and currently holds the position of a senior researcher at Department of Applied Radiochemistry. He works in the team of scientists headed by prof. B.Ya.Zilberman. His main interests are computer simulation of extraction, precipitation and evaporation equilibrium applied to Nuclear Fuel reprocessing. He has developed several software packages for simulation of stationary and dynamic component distribution through the stages of extraction units. He has publishes more than 20 papers and is working currently on his full doctor of science thesis.

A mathematical model has been proposed for calculating distribution coefficients of NPP spent fuel components in extraction systems with 30% TBP in a hydrocarbon diluent using the improved Rozen’s model based on the semi-empirical extraction equations. A computer code has been developed for simulation of the steady-state component concentration profiles throughout the extraction multistage unit, as well as the code for calculating the transient regime in the head and backwashing units of various PUREX process options for high burn-up spent fuel from nuclear power plants (NPP) with WWER (PWR) water-cooled reactors.\r\nThe analysis of published and newly obtained data on the influence of component concentration and temperature made possible to describe the equilibrium extraction of HNO3 and hexa- and tetravalent acti-nides as well as Zr and HTcO4 from nitric acid solutions into 30% TBP in the Purex process conditions in different extraction units except the ki-netics of ox-red reactions because of uncertainty of induction period.\r\nThe results of long-term bench experiments concerning transient regime of the head mixer–settler unit of the RT-1 complex were in good agreement with simulation results for entering the beyond-design-basis regime using the model of ideal mixing in a settling chamber, while the return from this regime is better described by plung-flow model. The effect of significant extraction of Pu at limiting loading of the solvent with uranium was confirmed allowing adequate description of the Pu accumulation in the uranium-free zone of the extraction– scrubbing unit. \r\nThe adequacy of the ideal mixing model for a unit of centrifugal con-tactors was demonstrated as applied to reprocessing of SNF from fast reactors. A separate experiment has shown that the Pu accumulation in the uranium-free zone does not occur in this case because of higher acidity of the aqueous phase and higher process temperature (so called IMPUREX process), while the Pu distribution between the contac-tor steps correlates with the U distribution.\r\nThe bench tests on transient regime in uranium backwashing unit have indicated on the applicability of the above model.\r\nThe drawbacks of Rozen’s model such as ignoring of hydrolyses and extraction of partially hydrolyzed or complexed species leading to the high number of empirical corrections, as well as impossibility to de-scribe salting-out effect and extraction by high concentration TBP urged us to look for a new approach to simulation of interphase equilibrium. The new “multireaction” model describes the extraction of actinides in various valent states, some fission products and admixed acids from nitric acid media in the presence of salting out agents by diluted TBP. It is based on a set of simultaneous chemical reactions of unlimited number of components, including salvation, hydrolysis, complexing and other reactions, characterized by apparent concentration constants.\r\n

Nickolay D. Goletskiy graduated from Physico-Chemical Department of SPb Institute of Technology (Technical Uni-versity), St-Petersburg, Russia, in 1998 and since then he has been working at Department of Applied Radiochemistry of Khlopin Radium Institute in the team headed by Prof. B. Ya. Zilberman. He received his Ph. D. in 2012 for thesis “Molybdenum extraction by TBP, HDBP, it’s zirconium salt and their mixtures from nitric acid media as applied to NPP SNF reprocessing” and holds now the position of leading researcher. His main interests include aqueous reprocessing of NPP SNF – process chemistry, mass transfer and control equipment, material balance of the radiochemical plant and closed fuel cycle as a whole. He is interested also in stoichiometry and kinetics of interphase extraction reactions, the use of simulation of extraction units and cycles. He supervises two applicants for Ph. D. and is the author of more than 50 papers and 19 patents.

NPP spent fuel reprocessing should be environmentally and commer-cially admissible and lead to recycling of regenerated material as well as to minimization of the radioactive waste final volume. The goals of the reprocessing, originally developed for Pu and U recovery for weapons material production now changed considerably on passing to NPP SNF, as the fuel fraction to be utilized (U and Pu) dropped to 95-97%, while the amount of non-utilized elements increased from 0.05% to 3-5% of the total SNF mass. Only < 5% of these elements are hazard-ous long-lived radionuclides, namely, α-nuclides (transplutonium elements, Np) and Tc, of which Np, Tc and Zr are notably extractable by diluted TBP causing difficulties in the 1st cycle of the PUREX process. \r\nZr, Tc and Np localization within the framework of the 1st extraction cycle of the Purex process is possible by several ways (See Table 1). Experi-mental data were obtained on Tc localization in the HL raffinate as applied to the RT-1 flowsheet, in the zirconium and tritium strip product, (Superpurex-1) and in the Pu and Np strip product or in the strip product from the unit of the barrier reductive scrubbing of the uranium extract as applied to six-block extraction scheme in the project of the Pilot Demonstration Center (PDC). All above variants were implemented at RT-1 plant or successively tested on “hot’ rigs for PDC using real WWER (PWR) fuel of high burn-up (40 -70 GW*d/t HM) and 2,5 – 6 y. storage. The obtained PF/DF for PDC flowsheets are shown in Table 2. \r\nThe possibilities of Np localization before and after Pu stripping with the opposite Tc withdrawal were also examined, and the first one (see Fig.) was proposed to be implemented at PDC for production of REMIX fuel made of dioxides of regenerated Pu and enriched regenerated U, suitable for uploading of 100% WWER-1000 active zone. At the same time the process provides the HLW evaporation without Zr/Mo and Ba(NO3)2 precipitate formation. The flowsheet should be tested using real spent fuel of PWR type. \r\nNo withdrawal of Pu, Np, Zr and Tc separately to generally different products, as well as Np with Tc together is possible in the frame of six-unit extraction flowsheet t excluding precipitation-free HLW evaporation which is the limiting factor for optimization of process chain as a whole.\r\n