Plutonium uranium oxide

The reactor is a 1000-MWe, sodium-cooled, mixed plutonium-uranium oxide fuel unit with a thermal rating of 2500 MW. Sodium at the reactor outlet temperature of 1100°F is used to produce steam at 3500 psia... 

MO fuel designed for use in existing LWRs is typically exposed to bum-ups greater than 40 x 10 MW-d/t. The discharged MO fuel has essentially the same uranium enrichment as uranium oxide fuel, but has a greater total amount of plutonium. 

One energy source that first appeared to be highly attractive was nuclear power. The problem with nuclear power is that some costs were hidden in its initial development. Especially pernicious is the disposal of uranium oxide fuel after it has become depleted. It can be reprocessed, but at considerable expense, and the product plutonium can be used for weapons. In the United States the plan is to bui y... 

Stanley JA, Edison AF, Mewhinney JA, et al. 1978. Inhalation toxicology of industrial plutonium and uranium oxide aerosols II. Deposition, retention and dosimetry. Health Phys 35(6) 888. 

The Purex process, ie, plutonium uranium reduction extraction, employs an organic phase consisting of 30 wt % TBP dissolved in a kerosene-type diluent. Purification and separation of U and Pu is achieved because of the extractability of U02+2 and Pu(IV) nitrates by TBP and the relative inextractability of Pu(III) and most fission product nitrates. Plutonium nitrate and U02(N03)2 are extracted into the organic phase by the formation of compounds, eg, Pu(N03)4 -2TBP. The plutonium is reduced to Pu(III) by treatment with ferrous sulfamate, hydrazine, or hydroxylamine and is transferred to the aqueous phase U remains in the organic phase. Further purification is achieved by oxidation of Pu(III) to Pu(IV) and re-extraction with TBP. The plutonium is transferred to an aqueous product. Plutonium recovery from the Purex process is ca 99.9 wt % (128). Decontamination factors are 106 — 10s (97,126,129). A flow sheet of the Purex process is shown in Figure 7. 

Tokai, K. Ooe, A. Manufacture of mixed oxide (MOX) pellets containg uranium oxide and plutonium oxide for fuel rods for power generation. JP 94-225519, Chem Abstr. 1996, 124, 272906. 

The projected installed nuclear electric capacity through the year 2000 is shown in Figure 1. The total installed nuclear capacity is expected to rise from 5.9 GW at the end of 1970 to 102 GW in 1980 and to 1200 GW in the year 2000. This represents 1.7%, 15%, and 54%, respectively, of the nation s total electric generating capacity in those years. In this projection, the greatest proportion of the nuclear capacity is contributed by LWRs. Plutonium, in the form of mixed plutonium and uranium oxide fuels, is recycled in LWRs beginning in 1977, and these mixed oxide fuels represent about 10% of all LWR fuels through the year 2000. HTGRs come on-stream in 1980. The LMFBRs go on-stream in 1987, and constitute almost 200,000 MW of installed capacity by the end of the century. 

The most important fuel for currently operated nuclear power stations (mainly light-water reactors) is - U-enriched uranium(IV) oxide. Also of importance are metallic uranium for the Magnox reactors and a few research reactors and uranium-plutonium mixed oxides for light-water reactors. Fuel production comprises extraction and dressing of uranium ores to uranium concentrates, conversion into UF, the uranium compound used for enrichment of the BSy.jjjotope, enrichment of and production of fuel from enriched UF5 (reconversion). 

Uranium-plutonium mixed oxides Uranium-plutonium mixed oxides (MOX) are becoming increasingly important, since plutonium is produced during the reprocessing of spent fuel elements. In these mixed oxide fuel elements a mixture of uranium(IV) and plutonium(IV) oxides with a plutonium content of 3 to 4% is utilized instead of ca. 4% 235u-enriched uranium(IV) oxide. Such fuel elements have similar nuclear physical properties to the standard elements with and can therefore be used in their place. 

In their manufacture uranium(IV) oxide is mixed with the appropriate quantity of plutonium(IV) oxide, the mixture pressed into pellets and then sintered (termed coprocessing in the USA). Uranium(IV) oxide is produced by one of the above-described processes and plutonium(IV) oxide from the aqueous nitrate solution produced during reprocessing by precipitating it as plutonium oxalate and calcining the oxalate. 

Reduction and oxidation (redox) steps are major process steps in the Purex process. Use is made of redox reactions to alter the valency of plutonium, uranium or neptunium with the object of producing these metals with a high degree of purity. 

Consider, for example, a PWR containing 80 te of uranium oxide enriched to 3%, with an average moderator temperature of 310 C. The effective specific volume of uranium oxide pellets will be 1.11 x 10 m /kg. The fission cross-section at a moderator temperature of 20°C/293 K, t/(293), is 582 x 10 m for U-235, while the corresponding figure for plutonium-239 is 743 x 10 m. Calculate NfOf at the start of life and at the end of life, and hence deduce the variation in the constant of proportionality, a. 

The bismuth phosphate process consisted of a number of steps in which plutonium is made alternatively soluble and insoluble. Fuel elements containing plutonium, uranium, and fission products were first dissolved in nitric acid. Plutonium was reduced to the tetravalent state by addition of sodium nitrite. Plutonium phosphate Pu3 (P04)4 was coprecipitated with bismuth phosphate BiP04, by addition of bismuth nitrate and sodium phosphate. Coprecipitation of uranium was prevented by the presence of sufficient sulfate ion to form anionic UO2(804)2. The BiP04 precipitate was redissolved in nitric acid and subjected to two decontamination cycles to purify the plutonium. In each cycle the plutonium was oxidized to the soluble hexavalent state by NaBiOs or other strong oxidant. Next bismuth phosphate was again precipitated, to remove fission products while hexavalent plutonium remained in solution. Then plutonium was reduced to the tetravalent state and again coprecipitated with bismuth phosphate. 

In this process, oxide fuel is dissolved in a molten chloride salt mixture through which Q2-HCI gas is flowing. Dissolved uranium and plutonium are then recovered as oxides by cathodic electrodeposition at 500 to 700°C. The process was demonstrated with kilogram quantities of irradiated fuel, with production of dense, crystalline UO2 or UO2-PUO2 reactor-grade material. Difficulties were experienced with process control, off-gas handling, electrolyte regeneration, and control of the plutonium/uranium ratio. Development has been discontinued. 

The fissile fractions in spent LWR fuel elements amounts to 0.9% U and 0.5—0.7% 239 + 24ipy gy recovering these and returning them to the LWR fuel cycle the demand for new uranium and enrichment services is reduced by 30%. The uranium recovered may either be re-enriched and used in normal uranium oxide fuel or bl ded with the plutonium recovered to form mixed oxide (MOX) fuel elements ( 21.1). MOX fuel can also be made from recovered plutonium and depleted uranium. MOX fuel elements for LWRs contain up to 5% Pu+ Pu. Many tons of plutonium have already been used as MOX fuel in LWRs. 

Metal-complexation/SFE using carbon dioxide has been successfully demonstrated for removal of lanthanides, actinides and various other fission products from solids and liquids (8-18), Direct dissolution of recalcitrant uranium oxides using nitric acid and metal-complexing agents in supercritical fluid carbon dioxide has also been reported (79-25). In this paper we explored supercritical fluid extraction of sorbed plutonium and americium from soil using common organophosphorus and beta-diketone complexants. We also qualitatively characterize actinide sorption to various soil fractions via use of sequential chemical extraction techniques. 

Some elements are not suitable for electrodeposition from aqueous solution as the metal. Among these are the radionuclides plutonium, uranium, and thorium, which are prepared for alpha-particle spectral analysis by deposition of oxides. Other metals, such as lead, can also be deposited as oxides under empirically derived conditions (Laitinen and Watkins 1975). 

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