Plutonium irradiated fuel

Canadian reactors (CANDU) are fueled using natural uranium. The discharged fuel contains small amounts of plutonium, but the fissile uranium content is below that of natural uranium. Therefore, the irradiated fuel is not normally considered a candidate for economic reprocessing. 

Irradiated Fuel A historically important and continuing mission at the Hanford site is to chemically process irradiated reactor fuel to recover and purify weapons-grade plutonium. Over the last 40 years, or so, several processes and plants— Bismuth Phosphate, REDOX, and PUREX—have been operated to accomplish this mission. Presently, only the Hanford PUREX Plant is operational, and although it has not been operated since the fall of 1972, it is scheduled to start up in the early 1980 s to process stored and currently produced Hanford -Reactor fuel. Of nine plutonium-production reactors built at the Hanford site, only the N-Reactor is still operating. 

The research programme of the European Institute for Transuranium Elements was, from its very beginning, devoted to both basic research on advanced plutonium containing fuel and to fundamental research on actinide elements. Non-fuel actinide research in Europe started more than 20 years ago with the reprocessing of irradiated actinide samples. Since the first isolation and purification of transplutonium elements, actinide research developed steadily in close contact and cooperation with specialised laboratories in Western Europe and in the United States. 

In the chemistry of the fuel cycle and reactor operations, one must deal with the chemical properties of the actinide elements, particularly uranium and plutonium and those of the fission products. In this section, we focus on the fission products and then chemistry. In Figures 16.2 and 16.3, we show the chemical composition and associated fission product activities in irradiated fuel. The fission products include the elements from zinc to dysprosium, with all periodic table groups being represented. 

The uranium and thorium ore concentrates received by fuel fabrication plants still contain a variety of impurities, some of which may be quite effective neutron absorbers. Such impurities must be almost completely removed if they are not seriously to impair reactor performance. The thermal neutron capture cross sections of the more important contaminants, along with some typical maximum concentrations acceptable for fuel fabrication, are given in Table 9. The removal of these unwanted elements may be effected either by precipitation and fractional crystallization methods, or by solvent extraction. The former methods have been historically important but have now been superseded by solvent extraction with TBP. The thorium or uranium salts so produced are then of sufficient purity to be accepted for fuel preparation or uranium enrichment. Solvent extraction by TBP also forms the basis of the Purex process for separating uranium and plutonium, and the Thorex process for separating uranium and thorium, in irradiated fuels. These processes and the principles of solvent extraction are described in more detail in Section 65.2.4, but the chemistry of U022+ and Th4+ extraction by TBP is considered here. 

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%. 

Nuclear material that can be used for the manufacture of nuclear explosive components without transmutation or further enrichment, such as plutonium containing less than 80% plutonium-238, uranium enriched to 20% uranium-235 and uranium-233 or more any chemical compound or mixture of the foregoing. Plutonium, uranium-233 and uranium enriched to less than 20% uranium-235 contained in irradiated fuel do not fall into this category. 

Plutonium, uranium-233 and enriched uranium contained in irradiated fuel. 

Contactors with low inventory of process solutions are also important when the material processed is valuable, such as the plutonium recovered from irradiated fuel. Low inventory is also important in maintaining a close accountability of the total inventory of fissionable material processed. 

The ability of diethyl ether to extract uranyl nitrate from aqueous solution has been known for a hundred years and was the method chosen by the Manhattan Project to purify the uranium used in the first nuclear chain reactors. This solvent has numerous disadvantages. It is very volatile, very flammable, and toxic, and it requires addition of sodium, aluminum, or calcium nitrate to the aqueous phase to enhance extractions. When solvent extraction was first applied to recovery of uranium and plutonium from irradiated fuel, other oxygenated solvents less volatile than diethyl ether that were first used were methyl isobutyl ketone, dibutyl... 

Although only small quantities of Pu are formed, its half-life of 86 years is long enough that Pu persists in plutonium recovered for recycle and is short enough that is the greatest contributor to the alpha activity of plutonium in irradiated fuel. Although the quantities and activities of 2.85-year Pu are relatively small, its decay daughter can... 

Seaborg and associates [LI] had found that tetravalent plutonium [Pu(IV)] could be coprecipitated from aqueous solution in good yield with insoluble bismuth phosphate BiP04, made by adding bismuth nitrate and sodium phosphate to an aqueous solution of plutonium nitrate. The bismuth phosphate process was developed at the Metallurgical Laboratory, demonstrated at the X-10 pilot plant at Oak Ridge National Laboratory in 1944, and put into operation for large-scale recovery of plutonium from irradiated fuel at Hanford in early 1945. 

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. 

In the Aquafluor process [G4] developed by the General Electric Company, most of the plutonium and fission products in irradiated light-water reactor (LWR) fuel are separated from uranium by aqueous solvent extraction and anion exchange. Final uranium separation and purification is by conversion of impure uranyl nitrate to UFg, followed by removal of small amounts of PuF , NpFg, and other volatile fluorides by adsorption on beds of NaF and Mgp2 and a final fractional distillation. A plant to process 1 MT/day of irradiated low-enriched uranium fuel was built at Morris, Illinois, but was never used for irradiated fuel because of inability to maintain on-stream, continuous operation even in runs on unirradiated fuel. The difficulties at the Morris plant are considered more the fault of design details than inherent in the process. They are attributed to the attempt to carry out aqueous primary decontamination, denitration, fluorination, and distillation of intensely radioactive materials in a close-coupled, continuous process, without adequate surge capacity between the different steps and without sufficient spare, readily maintainable equipment [G5, R8]. 

In all cases, however, dissolution of irradiated fuel in nitric acid leaves some plutonium associated with undissolved fission products. This plutonium can be leached from the residue with mixed nitric and hydrofluoric acids or with mixed nitric acid and ceric nitrate, Ce(N03)4 [U2]. Residue from irradiated mixed UO2-PUO2 fuel was 99.94 percent dissolved in 4 h by treatment with 4 M HNO3-O.5 M Ce(IV). Ceric nitrate is preferred to HF in the secondary dissolution step because cerium is already present as a fission product, and its addition does not complicate subsequent solvent extraction. Use of Ce(IV) in the primary dissolution step is undesirable because it would convert all plutonium to the less extractive hexavalent state and would volatilize much of the ruthenium as RUO4. 

Figure 10.29 shows the principal steps in applying the Purex process to irradiated LMFBR fuel, step 7 of Fig. 10.28. The flow scheme and the compositions and locations of solvent, scrubbing, and stripping streams have been taken from the process flow sheet of a 1978 Oak Ridge report [Oil] describing a planned experimental reprocessing facility designed for 0.5 MT of uranium-plutonium fuel or 0.2 MT of uranium-plutonium-thoiium fuel per day. As that report gave process flow rates only for the uranium-plutonium-thorium fuel. Fig. 10.29 does not give flow rates for the uranium-plutonium fuel of present interest. This flow sheet shows the codecontamination step, in which flssion products are separated from uranium and plutonium the partitioning step, which produces an aqueous stream of partially decontaminated...

This section describes processes for recovering neptunium from irradiated uranium. Neptunium is an example of one of the numerous elements in irradiated fuel that could be recovered as by-products of extraction of uranium and plutonium in the Purex process,... 

Starting in 1970 one further processing variant has been investigated— the extraction of plutonium by tricaprylamine dissolved in diethylbenzene (19). Since the irradiated uranium from CANDU reactors has a very low residual uranium-235 content, there is little incentive to recover it. The amine process offers the advantages of small size and a simple, one-cycle arrangement to give the desired decontamination. A bench-scale pilot unit has demonstrated satis ctory performance of the flowsheet, and it is the first time amine has been used to extract plutonium fiom dissolved irradiated fuel. 

In the past few years work has resumed on the development of the process for immobilization of wastes in glass to adapt it to the types of wastes now anticipated (83). Since it is not certain that Canadian irradiated fuel will be processed to recover plutonium, this program also is... 

Early Work. The irradiated fuel, upon discharge from the reactor, comprises the residual unbumt fuel, its protective cladding of magnesium alloy, zirconium or stainless steels, and fission products. The fission process yields over 70 fission product elements, while some of the excess neutrons produced from the fission reaction are captured by the uranium isotopes to yield a range of hew elements—neptunium, plutonium, americium, and curium. Neutrons are captured also by the cladding materials and yield a further variety of radioactive isotopes. To utilize the residual uranium and plutonium in further reactor cycles, it is necessary to remove the fission products and transuranic elements and it is usual to separate the uranium and plutonium this is the reprocessing operation. 

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