Spent Fuel Solutions

The norm ISO 18213 gives detailed recommendations concerning accountability tank installations and procedures for volume calibration and measurement (ISO 2007b). It reflects the results of extensive experience collected at a number of spent fuel reprocessing plants 

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Serrano-Purroy, D., Baron, P., Christiansen, B., Glatz, J.P., Madic, C., Malmbeck, R., Modolo, G. 2005. First demonstration of a centrifugal solvent extraction process for minor actinides from a concentrated spent fuel solution. Separation Science and Technology 45 157-162. 

Tracer techniques, for example, are used to obtain very small but representative and measurable samples of highly radioactive spent fuel solutions. One millilitre of the solution is then spiked with a known amount of uranium and plutonium tracer isotopes. A few microlitres of the spiked solution are dried and shipped to SAL. One to fifty nanograms of uranium or plutonium extracted from this tiny sample are sufficient for a complete analysis representing the composition of half a tonne of irradiated fuel with an accuracy of 0.3 to 0.5%. 

Since 1958, more than 20 nuclides of actinides ranging from neptunium to einsteinium were identified and prepared for tracer studies. From neutron-irradiated uranium samples 2 9Np was adjusted to the pentavalent state and separated by TBP extraction from perchloric acid media. Plutonium-239 was separated by TBP extraction from nitric acid solution followed by anion exchange in a system of Dowex-1 resin and nitric acid. Neptunium-237 was separated from a spent fuel solution of JRR-1 (Japan Research Reactor -1) using anion exchange and TBP extraction. The TBP extraction in the hydrochloric acid medium is a simple and effective technique to purify neptunium from plutonium contamination. On the other hand, both anion exchange and solvent extraction with HDEHP could be used to separate tracer scale plutonium from irradiated neptunium targets. 

A 0.1 ml aliquot of the dissolved spent fuel solution was diluted by adding 50 mL of pure water. The sample solution was heated and dried, and was dissolved again in 3 mL of 0.5 moEdm hydrochloric acid. The resulted hydrochloric dissolver solution was loaded onto resin (8 mL column volume). Tertiary pyridine resin of a non-support type was used as pre-filter anion exchange resin. After sample loading, the column was fed with 4 column volumes of 0.5 mol/dm hydrochloric acid solution. The resin was rinsed by pure water of 4 column volumes. Then. 1 mL of resin was taken from the column and rinsed again by pure water. After the rinse, the rinsing water and resin were encapsulated in a plastic vial. Furthermore, 3 column volumes of 1 mol/dm sodium hydroxide solution were fed to the resin. After the alkali rinse, the resin and eluted sodium hydroxide solution were analyzed by y-ray spectrometry to confirm the adsorbed nuclides. 

IAEA (Kuno et al. 1989), the spike is a mixture of uranyle and plutonium nitrate carefully dried in a 10 ml penicillin vial. For the assay of spent fuel solutions and MOX products with U/Pu ratios around 100 1, each vial contains about 50 mg of uranium-enriched just below 20% in and 2 mg of plutonium with a Pu abundance of 95% or more. Calculations (Laszlo et al. 1991) and experience show that these amounts are appropriate to reach relative uncertainties around 0.1% in a single measurement of samples carrying about 100 mg of depleted uranium and 1 mg of high burn plutonium with a Pu abundance of 50%. 

Neodymium has been considered as an alternative tracer (Debievre 1997), as it is a fission product present in measurable concentrations in spent fuel solutions and can serve as in situ spike. The accuracy of TIMS measurements is now so good that it could be practical to use enriched or as a tracer for the direct measurement of the total content of fissile material in a tank, skipping altogether the verification of the volume or mass of solution in the tank (Debievre and Perrin 1991). 

In the 1970s (Pella and von Baeckman 1969 von Baeckman 1971), an instrument was developed for the automatic dilution and L-X-ray fluorescence analysis (XRFA) of U and Pu in spent fuel solutions with a precision and accuracy of about 1%. 

Figure 10.42 Corrosion rate in dissolved fast reactor spent fuel solution and in simulated solution [89].

The spent fuel solution from the dissolver, adjusted to 2 molar nitric acid, flows into the middle of the first column. From there it flows downward countercurrent to a 5 percent tributyl phosphate/95 percent hydrocarbon solvent, which is introduced into the bottom section. The uranium and plutonium transfer to the solvent, leaving most of the fission products in the acid phase, which passes out the bottom of the column. The U/Pu solvent solution is scrubbed in the top half of the column with 3 molar nitric acid to wash out additional fission products. 

The licensing process consists of two steps construction and operating license that must be completed before fuel loading. Licensing covers radiological safety, environmental protection, and antitru,st considerations. Activities not defined as production or utilization of special nuclear material (SNM), use simple one-step. Materials Licenses, for the possession of radioactive materials. Examples are uranium mills, solution recovery plants, UO fabrication plants, interim spent fuel storage, and isotopic separation plants. 

Nuclear power produces spent fuel that contains radionuclides that will emit radiation for hundreds and thousands of years. At present, they are being stored underground indefinitely in heavy, shock-proof containers. These containers could be stolen or may corrode with time, or leak as a result of earthquakes and tremors. Transportation and reprocessing accidents could cause environmental contamination. One solution is for the United States to go to breeder reactors, as has been done in other countries, to reduce the level and amount of radioactive waste. 

A recent and extremely important development lies in the application of the technique of liquid extraction to metallurgical processes. The successful development of methods for the purification of uranium fuel and for the recovery of spent fuel elements in the nuclear power industry by extraction methods, mainly based on packed, including pulsed, columns as discussed in Section 13.5 has led to their application to other metallurgical processes. Of these, the recovery of copper from acid leach liquors and subsequent electro-winning from these liquors is the most extensive, although further applications to nickel and other metals are being developed. In many of these processes, some form of chemical complex is formed between the solute and the solvent so that the kinetics of the process become important. The extraction operation may be either a physical operation, as discussed previously, or a chemical operation. Chemical operations have been classified by Hanson(1) as follows ... 

Dissolution of the oxidized portion of the surface of the spent fuel, that is, the release of uranium(VI) to the solution. This is described by the following expression ... 

Fig- 2. Radioelement concentrations in solutions contacted with powdered spent fuel (UO2 bum-up 50 MW d/kg U -y-dose rate 10 Mrad/h solid surface/solution volume ratio 1000/m at 25 °C) after sequential filtration (filter pore size 450 nm - white bars filter pore size 1.8 nm -> grey bars) solutions consist of concentrated brine (5 mol/kg NaCl) and simulated granitic groundwater (I = 2.8 x 10 1 mol/L, pH 8) (Geckeis et at. 1998). 

Fig. 3, Evolution of Am(HI), Eu(III) and U concentrations with time in spent fuel pellet leaching experiments (leachate 5 mol/kg NaCl solution anaerobic conditions) radionuclides found in ultrafiltered samples (uf filter pore size 1.8 nm) arc considered as truly dissolved radionuclide concentrations found in filtered samples (f filter pore size 450 nm) are attributed to truly dissolved + colloidal species the grey shaded area marks the fraction of colloidal radioelement species in solution the black arrow indicates the pH increase in solution during the leaching experiment (Geckeis et al. 1998).

The spent fuel element is still mainly UO2 and is dissolved in aqueous nitric acid, which is oxidizing enough to take the uranium to the VI oxidation state as UC>22+(aq) and Pu to Pu4+(aq) (the uranyl ion U022+ can be regarded as hydrolyzed U6+ see Section 13.6). Treatment of the solution of uranyl and plutonium(IV) nitrates with either an iron(II) salt or SO2 will reduce all the Pu to Pu3+(aq), which is not extractable with TBP, but will leave the uranium(VI) untouched (see Exercise 15.5). The solution is then equilibrated with TBP (which is immiscible with water) or TBP in an alkane solvent. The U022+ forms a neutral complex containing both TBP and the nitrate ions, which axe present in large excess ... 

Elemental Compositions of various Spent Fuels and Liquid Concentrations in a Dissolved Solution... 

Diglycol amides. TODGA and other diglycol amides displayed an affinity toward Ca(II) and Sr(II) from 2-3 M HN03 solutions (422). Thereby, recovery of not only Ans-Lns but also Sr(II) from spent fuels is contemplated (279, 422, 423). The extracted complexes are represented as [Sr(N03)2L2(HN03)], where L = TODGA. 

Riddle, C.L., Baker, J.D., Law, J.D. et al. 2004. Development of a novel solvent for the simultaneous separation of strontium and cesium from dissolved spent nuclear fuel solutions. Americas Nuclear Energy Symposium (ANES 2004), Deauville Beach Resort, Miami Beach, FL, October 3-6. 

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