Fission product oxides

There are little data available which can be used to predict the rates of uptake of the different vaporized radioactive elements or oxides. Since such data are important to the application of any fallout prediction model based on kinetics, a program has been started at this laboratory to measure the rates of uptake of a selected group of fission-product oxides under conditions approximating those found in the cooling fireball. The data from these measurements will be useful, not only as input to fallout models, but also for discovering the mechanisms which govern the rates of uptake. 

Carbides can be converted to oxides and reduced as above. Metal alloy fuels are already in suitable form and can be processed directly. Calcium is used as the reducing agent as the free energy of Reaction 1 is -6.6 kcal/mole at 1000°C. The other fuel and fission product oxides are reduced more easily than Th02, thus the reduction should be complete once the thoria is reduced. 

Dissolution. The aluminum cladding was dissolved in a solution of mixed sodium nitrate and sodium hydroxide, and the undissolved uranium, thorium, and fission-product oxides were separated by filtration and centrifugation. The oxides, together with some adherent sodium hydroxide, sodium nitrate, and sodium aluminate, were dissolved in a mixture of boiling 13 M nitric acid, aluminum nitrate, and potassium fluoride. From 16 to 48 h were required for dissolution. After dissolution was complete, the composition of the solution was as given in the first column of Table 10.17. 

Table 11.7 shows the thermal conductivities of several solidification products. In Fig. 11.8 maximum temperature differences are plotted against the age of the waste assuming 20 w/o fission-product oxides in the solid. For the calculations typical canister diameters of about 250 and 500 mm and typical thermal conductivities of 0.25 (particulate calcine), 1.2 (glass), and 10 (vitromet) W/(m °C) have been used [E2]. 

Figure 3.14. Relative partial molar Gibbs free energy of oxygen of the fission product oxides and of UO2...

These qualitative assumptions are not contradictory to the results of solid-state solubility and phase diagram studies. According to these findings, the rare earth oxides are completely or largely miscible with UO2, whereas the alkaline earth oxides and zirconium oxide show a limited solubility and M0O2 is only soluble to a neglibly small extent (Kleykamp, 1985). However, results from studies performed on binary systems are not directly applicable to multi-component fuel - fission product oxide systems. In general, the maximum solubility of the individual fission product species in the fuel matrix will be reduced in the presence of additional constituents. 

Estimated steady-state concentration of fission-product oxides to be produced in the thorium oxide by irradiation at a flux of 5 X 10 neutrons/(cm )(sec) and continuous blanket processing on a 250-day cycle [155]. 

Uranium dioxide fuel is irradiated in a reactor for periods of one to two years to produce fission energy. Upon removal, the used or spent fuel contains a large inventory of fission products. These are largely contained in the oxide matrix and the sealed fuel tubing. 

The reaction of vinylogous amides, or ketoaldehydes, with hydroxylamine produced 4,5,6,7-tetrahydro-l,2-benzisoxazole. A side product is the 2,1-benzisoxazole (Scheme 173) (67AHC(8)277). The ring system can also be prepared by the reaction of cyclohexanone enamines with nitrile oxides (Scheme 173) (78S43, 74KGS901). Base treatment produced ring fission products and photolysis resulted in isomerization to benzoxazoles (76JOC13). 

The chemical identities of the fission products determine their subsequent redistribution, those elements which are in the gaseous state at the temperature of the operation migrating to the cooler exterior of the fuel rods, and die less voltile elements undergoing incorporation in the fuel rod in solid solution. Thus caesium and iodine migrate to the gas fill which sunounds the fuel rod, and elements such as the rare earths and zirconium are accommodated in solid solution in UO2 without significant migration along the fuel rod radius. Strontium and barium oxidize to form separate islands which can be seen under the microscope. 

I.J. Hastings, (ed.). Fission Product Behavior in Ceramic Oxide Fuel. Adv. in Ceramics 17, Amer. Ceram. Soc. (1986). 

Steam), and oi er input from CORCON. It contains a library of thermodynamic properties je energies from bich vapor pressures are calculated) for chemical species (mostly elements, oxides, and hydroxide that might be formed by fission products and other melt constituents. 

The behaviour of irradiated uranium has been studied mainly with respect to the release of fission products during oxidation at high temperatures The fission products most readily released to the gas phase are krypton, xenon, iodine, tellurium and ruthenium. The release can approach 80-100%. For ruthenium it is dependent upon the environment and only significant in the presence of oxygen to form volatile oxides of ruthenium. 

Studies of the influence of irradiation on the kinetics of oxidation have been confined to post-irradiation work. In general, prior irradiation increases reactivity, although there are considerable inconsistencies in the enhancements obtained The effects can be derived from an increased surface area associated with the swelling voids produced in the metal by the irradiation, and can also probably arise to a lesser extent from chemical effects of the fission products. 

When the Plutonium Project was established early in 1942, for the purpose of producing plutonium via the nuclear chain reaction in uranium in sufficient quantities for its use as a nuclear explosive, we were given the challenge of developing a chemical method for separating and isolating it from the uranium and fission products. We had already conceived the principle of the oxidation-reduction cycle, which became the basis for such a separations process. This principle applied to any process involving the use of a substance which carried plutonium in one of its oxidation states but not in another. By use of this... 

There is a scarcity of oxygen-transport data for oxygen-deficient actinide oxide systems. Because of this, our understanding and predictive capabilities of the effect of the defect solid state on the properties of reactor fuel systems, as well as on the chemical state of fission products in these systems, are limited. 

Three classes of carbamoylmethylphosphoryl extractants were studied for their ability to extract selected tri-, tetra-, and hexavalent actinides from nitric acid. The three extractants are dihexyl-N,N-diethylcarbamoylmethylphosphonate (DHDECMP), hexyl hexyl-N,N-diethylcarbamoylmethylphosphinate (HHDECMP), and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphos-phine oxide 0< >D[IB]CMP0. The above three extrac-trants were compared on the basis of nitric acid and extractant dependencies for Am(III), solubility of complexes on loading with Nd(III) and U(VI), and selectivity of actinide(III) over fission products. 

In broad terms, the following types of reactions are mediated by the homolytic fission products of water (formally, hydrogen, and hydroxyl radicals), and by molecular oxygen including its excited states—hydrolysis, elimination, oxidation, reduction, and cyclization. 

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