Molten plutonium fuels

The second reactor discussed in this chapter is the LAMPRE. This is a molten plutonium fueled reactor which is under development at the Los Alamos Scientific Laboratory. Although only in its beginning stages of development, it is conceived as a high temperature (6oO°C) fast breeder reactor utilizing plutonium as the fuel. 

Early experimental work in electrorefining at Los Alamos by Mullins et-all ) demonstrated that americium could be partitioned between molten plutonium and a molten NaCl-KCl salt containing Pu+3 ions, and Knighton et-al(8), working at ANL on molten salt separation processes for fuel reprocessing, demonstrated that americium could be extracted from Mg-Zn-Pu-Am alloys with immiscible molten magnesium chloride salts. Work... 

Plutonium nitride. Unlike the corresponding uranium-nitrogen system, only the one plutonium nitride PuN exists. It is prepared by heating plutonium hydride in nitrogen at 250 to 400 C, by reacting plutonium metal with a hydrogen-ammonia mixture at 600°C, or by direction reaction of molten plutonium with nitrogen at 1000 C. Plutonium nitride forms solid solutions with UN. However, because of the appreciable volatility and dissociation of PuN at temperatures at about 1600 C and above, the ternary (U, Pu)N is less attractive as a nuclear fuel than pure UN [K2, S4]. 

Distribution coefficients may be further modified and operating temperatures reduced by dissolving uranium fuel in a low-melting metal such as bismuth or zinc. Separation of uranium from fission products by liquid extraction between molten bismuth and fused chlorides was extensively studied at Brookhaven National Laboratory [D5] in connection with the liquid-metal fuel reactor (LMFR), which used a dilute solution of in bismuth as fuel. Extraction of fission products from molten plutonium by fused chlorides was studied at Los Alamos [L2] in connection with the LAMPRE reactor. 

Short-term tests indicated that the practical upper limit for tantalum as a container for uranium is about 1450°C. However, attack below this temperature is significant. A tantalum crucible with a wall thickness of 0.06 in. was completely corroded after 50 h at 1275°C. Other investigations have shown that tantalum is not attacked by uranium-magnesium and plutonium-magnesium alloys at 1150°C. Extensive tests on components for molten metal fuel reactors have demonstrated that tantalum is a satisfactory material for several thousand hours of service in liquid-metal... 

Molten Salt Reactor (MSR). The MSR [3] uses a liquid molten-fluoride salt as fuel and coolant. The uranium or plutonium fuel is dissolved in the molten salt. Two test reactors were built. In the 1950s, the Aircraft Reactor Experiment operated normally with molten salt exit temperatures of 815 C with peak operating temperatures up to 860 C and very low primary system pressures. Work continued on MSR technology for power applications until 1976. The reactor can be built in large sizes with passive safety systems. 

Plutonium is being investigated as an alternate fuel for the molten-salt reactor. Although it is too early to describe a plutonium-fueled reactor in detail, it is highly probable that a suitable PuFa-fueled reactor can be constructed and operated. 

It may be feasible to burn plutonium in molten fluoride-salt reactors. The solubility of PuFa in mixtures of LiF and BeFa is considerably less than that of UF4, but is reported to be over 0.2 mole % [8], which may be sufficient for criticality even in the presence of fission fragments and non-fissionable isotopes of plutonium but probably limits severely the amount of ThF4 that can be added to the fuel salt. This limitation, coupled with the condition that Pu is an inferior fuel in intermediate reactors, will result in a poor neutron economy in comparison with that of U -fueled reactors. However, the advantages of handling plutonium in a fluid fuel system may make the plutonium-fueled molten-salt reactor more desirable than other possible plutonium-burning systems. 

As a solvent for liquid-metal fuels, bismuth is a natural choice because it dissolves uranium and has a low cross section for thermal neutrons. As a result, research work at Brookhaven National Laboratory has centered on bismuth-uranium fuels. Other po.s.sible liPlutonium System (LAIMPRE) [14] and dispersions of uranium oxide in liquid metals, NaK [15] or bismuth [10]. The limited solubility, of uranium in bismuth is trouble.some in some designs. More concentrated fuels can be obtained by using slurries or dispersions of solid uranium compounds in bismuth. Among the. solids which have been suggested are intermetallic compounds [10] uranium oxide [10], uranium carbide, and uranium fluoride. Use of a dispersion avoids the limited concentration but introduces other problems of concentration control, stability, and erosion. 

Basic components. Before discussing the Los Alamos Molten Plutonium Reactor (LAMPRE) proposal in detail, the following resume will treat some of the possibilities for the three basic components of a power reactor the fuel, the container, and the coolant. 

Several components are required in the practical appHcation of nuclear reactors (1 5). The first and most vital component of a nuclear reactor is the fuel, which is usually uranium slightly enriched in uranium-235 [15117-96-1] to approximately 3%, in contrast to natural uranium which has 0.72% Less commonly, reactors are fueled with plutonium produced by neutron absorption in uranium-238 [24678-82-8]. Even more rare are reactors fueled with uranium-233 [13968-55-3] produced by neutron absorption in thorium-232 (see Nuclear reactors, nuclear fuel reserves). The chemical form of the reactor fuel typically is uranium dioxide, UO2, but uranium metal and other compounds have been used, including sulfates, siUcides, nitrates, carbides, and molten salts. 

In this process, uranium metal is electrodeposited at the cathode, while plutonium and other transuranium elements remain in the molten salt as trichlorides. Plutonium is reduced in a second step at a metallic cathode to produce Cd—Pu intermetallics. The refined plutonium and uranium metals can then be refabricated into metallic fuel (137). 

Recent studies on the electrochemical behavior of plutonium in molten salts have mainly been performed in LiCl— KCl based melts. The electrorefining step in a pyroprocessing procedure for the recycling of nuclear fuel from the Integral Fast Reactor (IFR) Program has been... 

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... 

Preliminary investigation has shown that most fission products are not soluable in alkali metal nitrate melts and that they are not dissolved by addition of 100% nitric acid vapor. If these characteristics are verified by further experiments, a fission product separation is easily envisioned. One could react the fuel with the molten nitrate, dissolve the uranate with the addition of 100% nitric acid, and separate the uranium from the remaining solids, which should consist of both plutonium dioxide and fission products. 

Larger samples are only partially chlorinated, since molten PuClj prevents the reaction going to completion [234], whereas plutonium(Hl) oxalate is completely converted into PuClj at 600-650 C, but the product is contaminated with carbon [234]. Plutonium(III) "carbonate", in contrast, is quantitatively converted between 4(X> and 550 C [234]. Phosgene has been used in a laboratory study upon the reprocessing (by chlorination) of neutron-irradiated uranium fuels [1491]. 

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. 

Pyrometallurgical processes investigated include slagging of molten irradiated uranium, plutonium extraction by silver, plutonium volatilization, and fused-salt extraction (78). Interest in these approaches ended with the selection of uranium dioxide as the CANDU fuel. 

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