Plutonium processing solution

Nuclear Waste Reprocessing. Liquid waste remaining from processing of spent reactor fuel for military plutonium production is typically acidic and contains substantial transuranic residues. The cleanup of such waste in 1996 is a higher priority than military plutonium processing. Cleanup requires removal of long-Hved actinides from nitric or hydrochloric acid solutions. The transuranium extraction (Tmex) process has been developed for... 

An overview is given of plutonium process chemistry used at the U. S. Department of Energy Hanford, Los Alamos National Laboratory, Rocky Flats, and Savannah River sites, with particular emphasis on solution chemistry involved in recovery, purification, and waste treatment operations. By extrapolating from the present system of processes, this paper also attempts to chart the future direction of plutonium process development and operation. Areas where a better understanding of basic plutonium chemistry will contribute to development of improved processing are indicated. 

The Los Alamos MPL can accept either plutonium nitrate solution or plutonium oxide as feed. If the feed is the nitrate solution, then the process steps are precipitation of... 

Spent Zirflex Process Decladdent. The basic perceived need is to devise an3 develop a simple process for selective and efficient removal of plutonium (and 21 1Am) from spent Zirflex process decladdent solution. To satisfy this need, it may be necessary--or prove beneficial—to determine, by appropriate physiochemical methods, the nature of the plutonium (and americium) species in the decladding solution. Availability of a satisfactory transuranium removal scheme may be one of the key factors in devising an alternative to storage in expensive double-shell tanks for spent Zirflex process solution at the Hanford site. 

Precipitation Processes. Plutonium peroxide precipitation is used at Rocky Flats to convert the purified plutonium nitrate solution to a solid (14) the plutonium peroxide is then calcined to Pu02 and sent to the reduction step. The chemistry of the plutonium peroxide precipitation process is being studied, as well as alternative precipitation processes such as oxalate, carbonate, fluoride, and thermal denitration. The latter method shows the most promise for cost and waste reduction. 

Purex [Plutonium and uranium recovery by extraction] A process for the solvent extraction of plutonium from solutions of uranium and fission products, obtained by dissolving spent nuclear fuel elements in nitric acid. The solvent is tri-n-butyl phosphate (TBP) in... 

Plutonium is subsequently stripped to an aqueous phase containing NH20H HN03 in the CC Column. In order to increase the plutonium concentration of the CC Column product, a portion of this stream (CAIS) is recycled to the CA Column after adjustment with HNO3. The remainder of the stream (CCP) is routed to the product concentrator. The resulting concentrated and purified plutonium nitrate solution is suitable feed to other processes for conversion to the desired product form (e.g., metal or plutonium dioxide). The remainder of the PRF solvent extraction system consists of a series of columns to wash the TBP-CCI4 solvent and prepare it for reuse. 

Purex [Plutonium and uranium recovery by extraction] A process for the solvent extraction of plutonium from solutions of uranium and fission products, obtained by dissolving spent nuclear fuel elements in nitric acid. The solvent is tri- -butyl phosphate (TBP) in kerosene. First operated by the U.S. Atomic Energy Commission at its Savannah River plant, SC, in 1954 and at Hanford, WA, in 1956. Now in operation, with modifications, in several countries. Sites include Savannah River (SC), Cap de la Hague (France), Marcoule (France), Sellafield (England), Karlsruhe (Germany), and Trombay (India). See also Recuplex. 

The redox methods are well known for the purification and concentration of plutonium from the Purex process solutions. 

The photochemical reduction of a solution containing both uranium(VI) and plutonium(IV) is also of interest for reprocessing applications. Early experiments (12a) showed a significant reduction of plutonium(IV) by light in Purex-type process solutions. Since the quantum yield for plutonium redox reactions is about one-tenth that for uranyl reduction (7b,c) the most likely path of plutonium(IV) reduction in these experiments appears to have been by uranium(IV) or uranium(V) generated by photochemical reduction of uranyl by other components of the solutions. Further experiments in this area would be useful. 

Tljie U-Pu separation is bas d on the much lower extractability of Pu3 ions by TBP than of Pu ions and the relative ease of oxidation and reduction of+plutonium in solutions. The original Pur x process utilized Fe2 to achieve the reduction of Pu1 to Pu. Since nitrite ions, whjch are generally present in nitric acid solutions, reoxidize Pu3 and thus affect the net reduction rate, a "holding reductant" is added to scavenge nitrite ions. Sulfamate ion, NH2S03 is an effective holding reductant and this led to the selection of ferrous sulfamate, Fe(NH2S03)2as the... 

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. 

PuOj from direct calcination of Pu(N03 )4. The precipitation steps of the above processes can be avoided by the direct calcination of the plutonium nitrate solution to PuOa. Calcination has been carried out at 350 C in a liquid-phase screw calciner. Half a mole of ammonium sulfate per mole of plutonium is added to the feed solution to increase the production of reactive PuOi. The calcination time and temperature must be low enough to minimize sintering, which would otherwise reduce the chemical reactivity of the oxide particles for subsequent conversion to a halide. 

A well-designed Purex plant aims for as complete recycle of solvent as possible, to minimize costs of solvent makeup and disposal. Solvent from the uranium purification section usually contains so few contaminants or degradation products that it can be reused a number of times without cleanup. On the other hand, solvent that has processed solutions containing hi activity of fission products and plutonium carries traces of these contaminants, uranium, nitric acid, dibutyl phosphate, and other radiolytic degradation products of TBP and dodecane. Uranium and plutonium should be recovered because of their value. Fission products should be removed to prevent product contamination in later cycles. Dibutyl phosphate should be removed because it forms strong complexes with tetravalent zirconium and plutonium that would impair ability of the solvent to reject zirconium and separate plutonium from uranium. 

Following the purification of the plutonium nitrate solution arising from reprocessing operations, the liquor enters the Pu finishing cycle. An identical process is used in both the Magnox and THORP reprocessing systems, which is as follows ... 

Critical thicknesses of infinite slabs of plutonium nitrate solution have been derived from data on spheres and cylinders, but no data appear in the open literature on the direct measurement of critical thicknesses for thin slabs of plutonium nitrate solution. Experiments have been conducted at the Critical Mass Laboratory of the Pacific Northwest Laboratory using a variable-thickness Blab-type vessel to determine critical thicknesses of bare and reflected infinite slabs of p lutonlum nitrate solution. The esmerlments provide date lor nuclear criticality safety applications in handling and processing plutonium and for checking computational methods. 

Certain operations in fissile material processing plants typically involve dissolution of uranium metal in nitric acid. To allow nuclear safety evalnatimm of such operations, a series of criticality calculatiohs was performed on systems of enriched uranium metal spheres in uranyl nitrate solution. Related calculations have been reMrted on plutonium metal in plutonium nitrate solution and on uranium metal slabs in uranium water mixtures. ... 

Experiments were recently completed to provide data on the effectiveness of soluble poisons on the criticality of plutonium-uranium solution mixtures. Although some data have been reported on Pu solutions, none have been available for U + Pu solutions. The current experiments provide data on systems typical of LMFBR fuel compositions. Gadolinium was chosen for this study because of its high neutron cross section, high solubility, and compatibility in the separation process i.e., gadolinium is more easily separable from Pu than cadmium or boron. The data can have applications in criticality control and prevention and serve also as experimental benchmarks for validating calculational techniques and cross-section sets. 

Data from critical experiments w required on the criticality of plutonium-uranium nitrate solutions to accurately establish criticality control limits for use in pro-cesang and handling of breeder-ty fueb. Since the flid must be processed both safely and economically, it b necessary that criticality considerations be based on accurate experimental data. Previous expieriments have been reported on plutonium ranium solutions with plutonium weight ratios extending up to some 38 wt% [plutonium weight... 

Experiences with aqueous chemistry and behavior of the transuranium elements obtained in nuclear fuel reprocessing and plutonium processing are only of limited relevance for PWR primary coolants with the extremely low concentrations of these elements in a boric acid—LiOH solution of varying composition. The plutonium polymers which are formed in less acid and neutral solutions and which have been reported to show the highest plate-out potential (e. g. Wilkins and Wisbey,... 

The most widely employed method for plutonium reprocessing used today in almost all of the world s reprocessing plants is the Purex (plutonium-uranium reduction extraction) process. Tributylphosphate (TBP) is used as the extraction agent for the separation of plutonium from uranium and fission products. In effecting a separation, advantage is taken of differences in the extractability of the various oxidation states and in the thermodynamics and kinetics of oxidation reduction of uranium, plutonium, and impurities. Various methods are in use for the conversion of plutonium nitrate solution, the final product from fuel reprocessing plants, to the metal. The reduction of plutonium halides with calcium proved to be the best method... 

On account of its large practical importance, polymer formation in hydrolyzed plutonium(iv) solutions has attracted much interest [203]. This polymer is formed fairly rapidly [204,205]. The reaction is faster, and more extensive, the higher the temperature. As long as ionic plutonium(iv) is present in detectable amounts, the rate of polymerization is proportional to the concentration of this component, and inversely proportional to the square of the acidity. When the ionic plutonium(iv) has been consumed, the rate depends in a rather complicated manner upon the concentration of other oxidation states present [205]. If the polymer is allowed to age, depolymerization becomes very slow even if the concentration of acid is fairly high [204]. The colloid behaves very differently from ionic plutonium(iv) in the extraction and ion-exchange procedures used in the processing of plutonium, and is also apt to transform into a precipitate. The conditions should therefore be chosen so that the formation of the colloid is... 

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