Thermal reactor fuels

Thermal oxide reprocessing plant, 6, 885 Thermal reactor fuels, 6,926 dissolution, 6,927 irradiated... 

Uranium tetrafluoride is a key intermediate in the production of thermal reactor fuels. It may be prepared directly from uranyl solutions by reduction of the U to U and addition of HF to precipitate UF4. A number of processes have been developed to produce UF4 by this wet route, which may be used to produce UF4 at the ore processing site. These employ iron, S02/Cu or electrolysis for the reduction step, the latter being preferred since it introduces no contaminants into the solution. The reduction of u in various media has been studied to assess the effect of complexation on the reduction reaction. The standard potential for the reduction of U02 to in 1 M HCIO4 has been given as +0.32 V. " The overall formation constants of fluoride complexes in 1 M NaCl were found to be log 82= 13.12, logj83 = 17.46 and log/84 = 21.8. Although wet processes have been developed as a short cut to UF4, the most widely used process at present involves dry processing. 

When more them one solute is involved in the consideration of the process design, the situation becomes much more complex since the extraction behaviours of the different solutes will usually be interdependent. In the case of irradiated thermal reactor fuels the solvent extraction process will be dealing with uranium containing up to ca. 4% of fission products and other actinides. These will have only a minor effect on uranium distribution so that a single-solute model may be adequate for process design. However, in some cases nitric acid extraction may compete with U02 extraction and a two-solute model may be needed. In the case of breeder reactor fuels the uranium may contain perhaps 20% of plutonium or thorium. Neptunium or protactinium levels in such fuels may also not be negligible and, under these circumstances, the single-solute... 

Solvent extraction processes (i) Thermal reactor fuels... 

In thermal reactors fueled with plutonium, the number of neutrons produced per neutron absorbed is less than 2.0 and breeding is impossible. For U, on the other hand, this number is substantially greater than 2.0, and breeding is practicable in a thermal reactor. In fast reactors, the number of neutrons produced per neutron absorbed is close to the total number of neutrons produced per fission, so that breeding is possible with both and plutonium. Breeding as here defined is not possible with U, because there is no naturally occurring isotope from which can be produced. 

Thermal reactor fuel preparation 65.2.3.4 Fast reactor fuel preparation... 

There are two breeder reactor fuel cycles. One involves the irradiation of U/ Pu oxide fuel with fast neutrons and is at the prototype stage of development. The other involves the irradiation of Th/ U oxide fuel with thermal neutrons and is at the experimental stage. Fuel from the U/ Pu cycle may be reprocessed using Purex technology adapted to accommodate the significant proportion of plutonium present in the fuel. Increased americium and neptunium levels will also arise compared with thermal reactor fuel. The Th/ U fuel may also be reprocessed using solvent extraction with TBP in the Thorex (Thorium Recovery by Extraction) process. In this case the extraction chemistry must also take account of the presence of Pa arising as shown in Scheme 2. 

The release is one of all gases and volatiles contained in about 2 kg of thermal reactor fuel and is only 3.5% of the activities listed in Table III as forming the inventory in fuel in which 1 MW of thermal power was generated during normal reactor operations. The absolute magnitude of the release... 

Fig. 6 Separation of lanthanide fission products present in dissolver solution of thermal reactor fuel. Column reverse phase Ci8, mobile phase camphor-10-sulfonic acid (0.05 M), and a-HIBA (0.1 M) pH 3.85, flow rate 1 ml/min. Postcolumn reaction with Arsenazo IB detection 655 nm.

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