Plutonium dissolution solution

The problem of recovering the plutonium contained in the Pu/Al target dissolution solutions is trivial in comparison with the difficulties discussed above. The strong affinity exhibited by tertiary amine nitrates (TLA or T0A) for Pu(IV) was exploited to develop the following processes ... 

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. ... 

Dissolution. Plutonium is solubilized in nitric acid solutions at Rocky Flats. The feed material consists of oxide, metal and glass, dissolution heels, incinerator ash and sand, slag, and crucible from reduction operations. The residues are contacted with 12M HNO3 containing CaF2 or HF to hasten dissolution. Following dissolution, aluminum nitrate is added to these solutions to complex the excess fluoride ion. 

The electrolyte salt must be processed to recover the ionic plutonium orginally added to the cell. This can be done by aqueous chemistry, typically by dissolution in a dilute sodium hydroxide solution with recovery of the contained plutonium as Pu(OH)3, or by pyrochemical techniques. The usual pyrochemical method is to contact the molten electrolyte salt with molten calcium, thereby reducing any PUCI3 to plutonium metal which is immiscible in the salt phase. The extraction crucible is maintained above the melting point of the contained salts to permit any fine droplets of plutonium in the salt to coalesce with the pool of metal formed beneath the salt phase. If the original ER electrolyte salt was eutectic NaCl-KCl a third "black salt" phase will be formed between the stripped electrolyte salt and the solidified metal button. This dark-blue phase can contain 10 wt. % of the plutonium originally present in the electrolyte salt plutonium in this phase can be recovered by an additional calcium extraction stepO ). 

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... 

Dissolution of a Primary Metal. This method is not likely to be used by many radioanalytical laboratories because of special facility requirements. Dedicated hoods and glove box facilities are used to handle the usual types and quantities of special nuclear materials. Samples of metallic plutonium, uranium, or neptunium can be purchased with certification of both purity and mass for use of the metal as the primary standard material. The metal is dissolved in acid and then diluted to prepare the solution of desired concentration. 

The flowsheet for the FFTF Pu02 production is shown in Figure 2. Briefly, the plutonium metal is converted to an impure oxide by burning the metal in air. This is followed by dissolution of the impure oxide in a 15.6 M HN03 - 0.5 M HF solution. The americium is separated from the plutonium by precipitation of the plutonium as the peroxide. Americium does not form an insoluble peroxide and stays in the filtrate with other cationic impurities. The active peroxide filtrate is slowly dripped into 9 M NaOH. The combination of strong alkali and heat destroys the peroxides and precipitates the americium as the hydroxide. Any residual plutonium in the filtrate, along with other cations, is precipitated also as the hydroxide. The flowsheet for the americium oxide production is shown in Figure 3. 

Other examples of redox-sensitive elements include heavy elements such as uranium, plutonium, and neptunium, all of which can exist in multiple oxidation states in natural waters. Redox conditions in natural waters are also indirectly important for solute species associated with redox-sensitive elements. For example, dissolution of iron (hydr)oxides under reducing conditions may lead to the solubilization and hence mobilization of associated solid phase species, e.g. arsenate, phosphate (see Sections 3.3.2.1, 3.3.3.2, and 3.3.4.1). 

The nitric acid solution from the dissolution of the fuel rod contents is filtered [poly(propene) fleece] or centrifuged, to remove suspended solids (zirconium- or molydenum- compounds and ruthenium and palladium alloys). The thus obtained fuel solution contains uranium, plutonium and the radioactive fission products. It is, after its composition is adjusted to the extraction conditions (3 molar in nitric acid and 240 to 300 g/L uranium) subjected to multi-cyclic extraction with tributylphosphate (dissolved in dodecane). Uranium and plutonium pass into the organic phase and are thereby separated from the fission products, which remain in the aqueous phase. 

Fluoride is known to separate zirconium fission product and plutonium from solvent degradation products. (5) Since fluoride is used to speed dissolution of thoria in nitric acid, and is already present in thorium solvent extraction process feed solutions, it was the first choice as an agent to use to improve thorium-DPB separation. 

Al, kk% Mg, 29% Pu, 2% Am, and other impurities. The residues are dissolved by placing the alloy metal in 0.35M HNO3 anc slowly adding concentrated HNO3. Time is allowed between additions of concentrated HNOj for the dissolution reaction to subside. The solution is adjusted to 0.1M F with concentrated HF to remove any polymeric or residual plutonium. After filtration through Whatman kl filter paper, the solution is adjusted to 7M HNO3. 

Objectives. The objectives of fuel dissolution are (1) to bring the uranium and plutonium in the fuel completely into aqueous solution (2) to complete the separation of fuel from cladding (3) to determine as accurately as possible the amounts of uranium and plutonium charged to reprocessing and (4) to convert uranium, plutonium, and fission products into the chemical states most favorable for their subsequent separation. 

An important objective of dissolution and the preconditioning of feed solution prior to extraction is to convert these fission-product elements into states that will not contaminate uranium, plutonium, or solvent in subsequent solvent extraction. 

The rate of dissolution of PUO2 in nitric acid is slower than UO2 and depends on the plutonium/uranium ratio, the methods used to fabricate fuel, and the conditions of irradiation. At one extreme, plutonium produced at low concentration in UO2 by transmutation dissolves almost as rapidly as the associated UO2. At the other extreme, plutonium present as PUO2 mixed mechanically with UO2 without proper sintering dissolves much more slowly and less completely than UO2. Plutonium present as a solid solution (U,Pu)02 at the concentration of 20 to 25 percent used in breeder-reactor fuel dissolves at an intermediate rate. ... 

Dissolution, described in Sec. 4.4, produces an aqueous solution of uranyl nitrate, plutonium(IV) nitrate, nitric acid, small concentrations of neptunium, americium, and curium nitrates, and almost all of the nonvolatile fission products in the fuel. With fuel cooled 150 days after bumup of 33,000 MWd/MT, the fission-product concentration is around 1700 Ci/liter. The fint step in the solvent extraction portion of the Purex process is primary decontamination, in which from 99 to 99.9 percent of these fission products are separated from the uranium and plutonium. Early removal of the fission products reduces the amount of required shielding, simplifies maintenance, and facilitates later process operations by reducing solvent degradation from radiolysis. 

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