Decontamination americium

Zr-Al coprocess waste test, the feed, extractant, and scrub flows were 1, 0.5, and 0.1 mL/min, respectively. For the high sodium concentration waste, the feed, extractant, and scrub flows were 0.75, 1, and 0.25 mL/min, respectively. Samples of raffinate were drawn for analytical analysis approximately five hours after equilibrium had been reached. The resultant decontamination factors agreed reasonably well with our calculations. For the coprocess waste run, we expected an americium decontamination factor of 200. We purposely built in a large, overkillM in the sodium waste run by increasing the organic to aqueous flow rates. The sodium waste run produced a raffinate that, when calcined, would be well below the guideline for alpha-free waste with no allowance for decay. Analytical analysis of feeds and raffinates confirmed our batch results in that actinides were fractionated from major waste constituents such as aluminum, zirconium, sodium, and fluoride. 

Berry JR, McMurray BJ, Jech JJ, et al. 1983. 1976 Hanford americium exposure incident Decontamination and treatment facility. Health Phys 45(4) 883-892. 

Nenot JC, Morin M, Lafurna J. 1971. Etude experimentale de la decontamination du squelette apres inhalation de nitrate d americium. Health Phys 21 395-400. 

EDTA which cloggs filters and, sometimes, even the columns. The feed pump is actuated by the signal of the gamma detector placed near the waste outlet pipe. This allows continuous feed of the active solution without the risk of polluting the waste with americium 241. On the POX. 11 column packed with 9 kg of stationary phase, one cycle served to extract 15.2 g of americium 241 present initially in 685 L of solution. The americium 241 concentration factor was 34, and the americium 241 waste decontamination factor was 250. The americium eluate contained only very small amounts of Fe and Cd. Decontamination factor of Am from Fe and Cd were 130 and 140, respectively. 

The development of the program for the production of transplutonium elements, americium 241, americium 243 and curium 244 in France required a major effort from the technological and chemical standpoints. Pre-existing hot cells were reconditioned and others were specially built for these production operations. From the chemical standpoint, the development of extractive chromatography on the preparative scale has allowed the definition of simple processes whose performance characteristics in our operating conditions have proved to be better than those obtained by liquid-liquid extraction. This type of process, initially developed for the treatment of Pu/Al targets, is ideal for the treatment of industrial wastes for their decontamination and for the production of americium 241. 

The process operated successfully in the plant. More than 98% of the americium was recovered from the cation exchange column as an acidic nitrate solution. Substantial quantities of sodium, iron, nickel, sulfate, and phosphate were removed. Decontamination from these impurities was satisfactory, but almost all the chromium and small amounts of nickel, iron, and lead remained. 

The DBBP extraction scheme provides excellent decontamination of plutonium and americium from all the other metals in the neutralized CAW solution. Kingsley(7) reports that distribution ratios for Fe3+, Al3+, Ca2+, and Mg between 30 vol% DBBP-CCI4 and neutralized CAW are, respectively, 0.11, <0.003, 0.025, and <0.0005. Primarily because of entrainment but partly because of extraction, small amounts of aluminum, iron, and sodium accompany 21tlAm into the dilute HNO3 strip solution. Richardson(9 ) in nonradioactive tests of the countercurrent DBBP 2 1Am recovery process observed decontamination factors in the range 80-180 for iron and in the range 2 x 103 - 1.5 x 105 for aluminum. 

The countercurrent DBBP z 1hn extraction process was operated on a plant-scale for about six years. During that time, it provided excellent recovery ( vl00%) of soluble plutonium in the feed and adequate decontamination of 21flAm from plutonium and other metallic impurities. The Am/Pu ratio in the WS-1 Column (Am strip column) product was 2.5/1 compared to only 1/1 for the batch extraction process americium product. Concentrations of calcium, aluminum, and other metallic impurities in the counter-current Am product stream were also much lower than in the Am product from the batch extraction process. 

Of particular interest is the possibility of decontaminating actinides from impurities by washing the loaded column with dilute acid or monovalent salts to elute the divalent impurities before stripping the column with 7M HNO3. This method is presently being evaluated for lead removal from americium using Bio-Rad AG MP-50. 

Table IV shows the results of three laboratory runs of the solvent extraction step (after the plutonium was removed by anion exchange). The major elements are shown before solvent extraction, after solvent extraction, in the 7M HNO wash, and in the final strip product. The americium remaining in the organic after stripping is also shown. Although there were some analytical discrepancies, the data show that americium was effectively recovered (except in Test 1, for which we have no explanation). Americium was decontaminated from aluminum and magnesium in all three runs.

Results of the pilot plant test of the 1iquid-1iquid solvent extraction test are shown in Table V. Only k Z of approximately 17 of americium strip product were used in the precipitation step. Decontamination from A1 and Mg was excel-... 

Analysis of the final Am02 product is shown in Table VI. A1 and Mg were below the detectable limits for these elements. The product met specifications of >95% Am02 with less than 0.5% Pu and less than 1% of any other single contaminant. The results of the preliminary lab scale extraction chromatography tests are shown in Table VII. Again, in spite of some analytical problems, it is evident that americium was decontaminated from aluminum and magnesium. A 7M HNO3 wash step is assumed to account for the americium loss. 

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. 

Synthetic waste solutions which simulate those often encountered in the PUREX process were tested in a SLM system to determine the efficiency for decontamination by removal of transuranic elements, such as americium (5). Diethyl N,N-... 

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