Uranium-plutonium oxide fuel

Program direction will include the study of cross-cutting efforts in both countries dealing with reactor safety, safety issues involving transportation, plutonium vitrification, treatment options, and mixed uranium-plutonium oxide fuel fabrication. The program will also focus attention on safety in the storage, handling, treatment, and disposition of fissile weapons materials. 

The second reactor reference design is a large sodium-cooled reactor with a power output capability between 500 and 1500 MWe that utilizes uranium-plutonium oxide fuel. This reactor design is supported by a fuel cycle based on an advanced aqueous process that would include a centrally located processing facility supporting a number of reactors. 

Testing of FAs with mixed uranium-plutonium oxide fuel and other types of nuclear fuel. 

All characteristics are presented for the option of uranium-plutonium oxide fuel. 

A large size (600—1500 MWei) loop-type reactor with mixed uranium—plutonium oxide fuel and potentially minor actinides, supported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors ... 

A variant of the HWR is the Eugen reactor developed by Japan. This reactor is heavy water-moderated but lightwater-cooled. It is fueled by mixed uranium—plutonium oxides. 

Salt Transport Processing (8, 9, 10, 11) The selective transfer of spent fuel constitutents between liquid metals and/or molten salts is being studied for both thorium-uranium and uranium-plutonium oxide and metal fuels. The chemical basis for the separation is the selective partitioning of actinide and fission-product elements between molten salt and liquid alloy phases as determined by the values of the standard free energy of formation of the chlorides of actinide elements and the fission products. Elements to be partitioned are dissolved in one alloy (the donor... 

Maximum power output per site Lower natural uranium consumption per MWh Reduced long-lived high-level waste generation during operation Capability to use mixed uranium and plutonium oxide fuel from recycled used fuel Minimal quantities of high-activity metal structure... 

The technical feasibility of the use of MOX (uranium + plutonium oxide) and UOX (uranium oxide) fuels for the recycling of minor actinides in fast reactors has been proven to some extent. Studies are still in progress for recycling in Light Water Reactors (LWR). 

Fuel type Cylindrical fuel elements with fuel pellets made of uranium-plutonium oxide, nitride or metallic fuel in steel claddings ... 

The CEFR is a sodium cooled, bottom supported 65 MW(th) experimental fast reactor fuelled with mixed uranium-plutonium oxide (the first core, however, will be loaded with uranium oxide fuel). Fuel cladding and reactor block structural materials are made of Cr-Ni austenitic stainless steel. It is a pool type reactor with two main pumps, and two loops for the primary and secondary circuit, respectively. The water-steam tertiary circuit has also two loops, with the superheated steam collected into one pipe that is connected with the turbine. CEFR s has a natural circuit decay heat removal system. 

By contrast, uranium fuels for lightwater reactors fall between these extremes. A typical pressurized water reactor (PWR) fuel element begins life at an enrichment of about 3.2% and is discharged at a bum-up of about 30 x 10 MW-d/t, at which time it contains about 0.8 wt % and about 1.0 wt % total plutonium. Boiling water reactor (BWR) fuel is lower in both initial enrichment and bum-up. The uranium in LWR fuel is present as oxide pellets, clad in zirconium alloy tubes about 4.6 m long. The tubes are assembled in arrays that are held in place by spacers and end-fittings. 

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. 

Uranium and mixed uranium—plutonium nitrides have a potential use as nuclear fuels for lead cooled fast reactors (136—139). Reactors of this type have been proposed for use ia deep-sea research vehicles (136). However, similar to the oxides, ia order for these materials to be useful as fuels, the nitrides must have an appropriate size and shape, ie, spheres. Microspheres of uranium nitrides have been fabricated by internal gelation and carbothermic reduction (140,141). Another use for uranium nitrides is as a catalyst for the cracking of NH at 550°C, which results ia high yields of H2 (142). 

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... 

Uranium is used as the primai-y source of nuclear energy in a nuclear reactor, although one-third to one-half of the power will be produced from plutonium before the power plant is refueled. Plutonium is created during the uranium fission cycle, and after being created will also fission, contributing heat to make steam in the nuclear power plant. These two nuclear fuels are discussed separately in order to explore their similarities and differences. Mixed oxide fuel, a combination of uranium and recovered plutonium, also has limited application in nuclear fuel, and will be briefly discussed. 

CSC atomization was developed by AEA Harwell Laboratories in the UK in the early 1970 s. Initially, the CSC process was used for the atomization of refractory and oxide materials such as alumina, plutonium oxides, and uranium monocarbide in nuclear fuel applications. Since it is well-suited to the atomization of reactive metals/alloys or those subject to segregation, the CSC process has been applied to a variety of materials such as iron, cobalt, nickel, and titanium alloys and many refractory metals. The process also has potential to scale up to a continuous process. 

In a typical fast breeder nuclear reactor, most of the fuel is 238U (90 to 93%). The remainder of the fuel is in the form of fissile isotopes, which sustain the fission process. The majority of these fissile isotopes are in the form of 239Pu and 241Pu, although a small portion of 235U can also be present. Because the fast breeder converts die fertile isotope 238 U into the fissile isotope 239Pu, no enrichment plant is necessary. The fast breeder serves as its own enrichment plant. The need for electricity for supplemental uses in the fuel cycle process is thus reduced. Several of the early hquid-metal-cooled fast reactors used plutonium fuels. The reactor Clementine, first operated in the Unired States in 1949. utilized plutonium metal, as did the BR-1 and BR.-2 reactors in the former Soviet Union in 1955 and 1956, respectively. The BR-5 in the former Soviet Union, put into operation in 1959. utilized plutonium oxide and carbide. The reactor Rapsodie first operated in France in 1967 utilized uranium and plutonium oxides. 

The Purex process is used for almost all fuel reprocessing today. Irradiated UO2 fuel is dissolved in HNO3 with the uranium being oxidized to U02(N03)2 and the plutonium oxidized to Pu(NC>3)4. A solution of TBP in a high-boiling hydrocarbon, such as n-dodecane, is used to selectively extract the hexavalent U02(N03)2 and the tetravalent Pu(NC>3)4 from the other actinides and fission products in the aqueous phase. The overall reactions are... 

Distribution ratios and transport were carried out on real HAW arising from dissolution of a mixed oxide of uranium and plutonium (MOX) fuel (burnup 34,650 MW d/tU), where uranium and plutonium have been previously extracted by TBP.86 The experiments were performed in the CARMEN hot cell of CEA Fontenay aux Roses with two dialkoxy-calix[4]arene-crown-6 derivatives (diisopropoxy and dini-trophenyl-octyloxy). High cesium distribution ratios were obtained (higher than 50) by contacting the HAW solution with diisopropoxy calix[4]arene-crown-6 (0.1 M in NPHE). Moreover, the high selectivity observed with the simulated waste was confirmed for most of the elements and radionuclides (actinides or fission products Eu, Sb, Ce, Mo, Zr, and Nd). The residual concentration or activity of elements, other than cesium, was less than 1% in the stripping solution, except for iron (2%) and ruthenium (8%) the extraction of these two cations, probably under a complexed... 

Since publication of this work, Japanese researchers have undertaken an effort to demonstrate the feasibility of direct dissolution of U02 from spent nuclear fuels by the TBP-HN03 complex in SC-C02.49 Ultimately, the project is directed at the extraction of both uranium and plutonium from mixed oxide fuels and from irradiated nuclear fuel. Ideally, soluble uranyl and plutonium nitrate complexes will form and dissolve in the C02 phase, leaving the FPs as unwanted solids. As in the conventional... 

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