SIMSISAK–a Method to Model Nuclide Transport in the SISAK System

A computer model that calculates the transport yield of a nuclide through an arbitrary SISAK experimental set-up has been developed. The model is intended to be used for two types of calculations connected to chemical studies of the heaviest elements. If the production cross section and the nuclide half-life are known, it can be used to estimate the number of decay events to be expected at the detection site. Consequently, if the number of atoms decaying in the detection cells is known, it can be used to estimate the production cross section or the half-life, provided that one of these properties is known

Homogenous recycling of transuranium elements from irradiated fast reactor fuel by the EURO-GANEX solvent extraction process

The EURO-GANEX process was developed forco-separating transuranium elements from irradiatednuclear fuels. A hot flow-sheet trial was performed in acounter-current centrifugal contactor setup, using a genuinehigh active feed solution. Irradiated mixed (carbide,nitride) U80Pu20 fast reactor fuel containing 20 % Pu wasthermally treated to oxidise it to the oxide form which wasthen dissolved in HNO3. From this solution uranium wasseparated to >99.9 % in a primary solvent extraction cycleusing 1.0 mol/L DEHiBA (N,N-di(2-ethylhexyl)isobutyramidein TPH (hydrogenated tetrapropene) as the organicphase. The raffinate solution from this process, containing10 g/L Pu, was further processed in a second cycle of solventextraction. In this EURO-GANEX flow-sheet, TRU andfission product lanthanides were firstly co-extracted intoa solvent composed of 0.2 mol/L TODGA (N,N,N′,N′-tetran-octyl diglycolamide) and 0.5 mol/L DMDOHEMA (N,N′-dimethyl-N,N′-dioctyl-2-(2-hexyloxy-ethyl) malonamide)dissolved in Exxsol D80, separating them from most otherfission and corrosion products. Subsequently, the TRUwere selectively stripped from the collected loaded solventusing a solution containing 0.055 mol/L SO3-Ph-BTP(2,6-bis(5,6-di(3-sulphophenyl)-1,2,4-triazin-3-yl)pyridinetetrasodium salt) and 1 mol/L AHA (acetohydroxamicacid) in 0.5 mol/L HNO3; lanthanides were finally strippedusing 0.01 mol/L HNO3. Approximately 99.9 % of the TRUand less than 0.1 % of the lanthanides were found in theproduct solution, which also contained the major fractionsof Zr and Mo

Evaluation of Thermochemical and Electrochemical Data for the Pyrochemical Partitioning Process.

Abstract not availableJRC.E-Institute for Transuranium Elements (Karlsruhe

Towards a DIAMEX process using high active concentrate. Production of genuine solutions

The efficiency of Minor Actinides (MA) recovery in the DIAMEX process has already been demonstrated using High Active Raffinate (HAR). The next step aims at the demonstration of reprocessing from High Active Concentrate (HAC) as feed, in view of an industrial application. The volume reduction would reduce the size of the installation to be used and thereby the costs of the process. The first step towards the demonstration of a DIAMEX process using HAC is the production of the genuine solutions. In the hot cell facility of ITU (Institute for Transuranium Elements), a HAR solution has been prepared, from small scale PUREX reprocessing of MOX fuel, and successfully subjected to a concentration/denitration process to obtain HAC. A final concentration factor (CF) of about 10 and an acidity of 4 M were reached. In the experiment a precipitate mainly composed of Sr, Zr, Mo, Sn and Ba was formed. MA precipitation was not significant (< 0.001%)

Flow-Sheet Design for an Innovative SANEX Process Using TODGA and SO3-Ph-BTP

AbstractAn innovative SANEX process was designed using an organic phase comprising TODGA in TPH + 5% 1-octanol and an aqueous phase containing SO3-Ph-BTP for the selective strip. The flow-sheet was optimized using batch data, single stage data and data from previously run TODGA processes. Recoveries for some key elements were calculated for selective strip section of the process using the SX Process program. According to the calculations, the An(III) are recovered from the PUREX raffinate with insignificant amounts of impurities. The presented flow-sheet will be used for a centrifugal contactor demonstration process which is expected to be successful

Investigation of the radiolytic stability of a CyMe4-BTBP based SANEX solvent

The radiolytic degradation of the 6,6'-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo[1,2,4]triazin-3-yl )-[2,2']-bipyridine (CyMe4-BTBP) based SANEX (selective actinide extraction) solvent has been investigated. As the solvent used in the extraction process is designed to separate trivalent actinides from lanthanides, the radiolytic degradation is mainly due to alpha decay of extracted minor actinide isotopes. A calculation of dose-rates was done by estimating the concentration of minor actinides in the solvent by fuel burn-up calculations and assumptions on dilutions in the subsequent reprocessing steps. The calculations showed that the main isotopes responsible for the dose-rate are Cm-242, Cm-244 and Am-241. Cm-242 is short-lived and has an impact only at short cooling times before reprocessing of the spent fuel. The dose-rates to a SANEX solvent in the reprocessing of standard spent LWR fuels are burn-up dependent and range from at least 0.03-0.2 kGy/h for UO2 fuels and from 0.4 to 0.8 kGy/h for MOX fuels. Fast reactor fuels yield dose-rates over 1 kGy/h. Based on these results, several radiolysis experiments were carried out in order to compare the effect of low LET external gamma radiation (0.2 kGy/h) and internal alpha radiation with different dose-rates (0.05, 0.2 and 1.0 kGy/h). Significant radiolytic degradation was shown in the gamma radiolysis and in the alpha radiolysis experiment at a dose-rate of 1 kGy/h. These experiments were continued up to an absorbed dose similar to 1200 kGy and >300 kGy, respectively. Comparing the alpha radiolysis results for 0.2 kGy/h and 1.0 kGy/h, up to an absorbed dose of similar to 120 kGy, no significant difference in the degradation for the different dose rates could be seen. The radiolytic degradation rate for gamma radiation was 40% higher than for alpha radiation