Authors:K. Gupta, S. Misra, S. Tripathi, and Manmohan Kumar
2-Ethylhexyl-2-ethylhexyl phosphonic acid (PC-88A) and Tributylphosphate (TBP) extractants have been attached to polypropylene
(PP) in granular, film and non-woven fabric forms, by a simultaneous γ-ray irradiation method. The extraction of plutonium
from the acidic radioactive liquid waste by modified polymers was studied by varying the γ-ray dose. The uptake of plutonium
was also studied by polyethylene (PE) in film form. This modified polymer also showed extraction capability for plutonium
from nitric acid medium. The uptake of plutonium depends upon the γ-ray dose as well as the nature and source of the polymer.
Liquid–solid extraction studies showed that the equivalent amount of uptake of plutonium on TBP–polyethylene film requires
twice the γ-ray dose as compare to TBP–polypropylene film. It was observed that at given γ-ray dose polypropylene fabric is
not sturdy, compare to the granules and films, and material leach out in aqueous phase. The presence of other solvents like
di-methyl formamide (DMF) and cyclohexane during γ-ray irradiation were able to enhance the extraction capabilities. The optimum
conditions established during this study was successfully applied for the separation of plutonium, uranium and thorium from
the fission products in acidic waste solution.
A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor
was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied.
The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was
studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate
was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.
Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium
reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin
(HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium
oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium
from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium
is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase
by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP
and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping
stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product
obtained can be handled on table top.
Authors:Jinping Liu, Hui He, Hongbin Tang, and Yanxin Chen
Both single stage and multi-stages experiments on stripping plutonium with N,N-dimethylhydroxylamine (DMHAN) as reductant with methylhydrozine (MMH) as supporting reductant were carried out. The effect
of contact time, temperature, acidity, concentration of DMHAN on back-extraction rate of plutonium was investigated in the
single stage experiment. The results demonstrated that the reaction of stripping Pu(IV) in the organic phase (30% TBP–kerosene)
1BF solutions by DMHAN exhibits excellent stripping efficiency. Under the given conditions, the back-extraction rate of plutonium
reaches 90% within 2 min. Higher temperature, lower acidity and the increased concentration of DMHAN benifit the stripping
reaction. The concentration profile of HNO3, uranium and plutonium were determined in a multi-stages mixer-settler after the steady state of the back-extraction, and
the multi-stages results show that the plutonium can be separated effectively from uranium. The recovery of plutonium and
uranium reach 99.995% or over 99.99% respectively. The separation factor of U from Pu (SFPu/U) is about 2 × 104.
Authors:Terry Hamilton, Doug Dasher, Tom Brown, Roger Martinelli, Alfredo Marchetti, and Steven Kehl
Plutonium-239 (239Pu) and plutonium-240 (240Pu) activity concentrations and 240Pu/239Pu atom ratios are reported for Brown Algae (Fucus distichus) collected from the littoral zone of Amchitka Island (Alaska), and at a control site at Unalaska, Alaska. The average 240Pu/239Pu atom ratio observed in dried F. distichus collected from Amchitka Island was 0.227 ± 0.007 (N = 5) and compares with the expected 240Pu/239Pu atom ratio in integrated worldwide fallout deposition in the Northern Hemisphere of 0.1805 ± 0.0057. In the absence of
any evidence of a local source of plutonium containing an elevated 240Pu/239Pu isotopic signature, the characteristically high 240Pu/239Pu content of F. distichus supports the view of the existence of a discernible, basin-wide non-fallout source of plutonium entering the subarctic Pacific.
Authors:S. Lee, P. Povinec, J. Gastaud, J. La Rosa, E. Wyse, and L. Fifield
Analysis of plutonium isotopes by Semiconductor Alpha Spectrometry (SAS), ICP-sector field mass spectrometry (ICP-MS) and
Accelerator Mass Spectrometry (AMS) was carried out in seawater samples collected from the Northeast Atlantic Ocean (nuclear
waste dumping sites) and Northwest Pacific Ocean. No particularly elevated levels of the atom ratios of 240Pu/239Pu compared to global fallout ratio (0.18) were found in the Northeast Atlantic Ocean seawater samples. The higher levels
of atom ratios of 240Pu/239Pu were found in the Northwest Pacific Ocean. This is mainly due to contribution from the local fallout from nuclear weapon
tests carried out at the Pacific Proving Grounds at the Marshall Islands.
The effect of physicochemical parameters such as pH, salinity (e.g. [NaCl]) and competitive cation (e.g. Ca2+ and Fe3+) concentration on the separation recovery of plutonium and uranium from aqueous solutions by cation exchange has been investigated.
The investigation was performed to evaluate the applicability of cation exchange as separation and pre-concentration method
prior to the radiometric analysis of uranium and plutonium isotopes in natural water samples. Application of the method to
test solutions of constant radionuclide concentration and variable composition (0.1, 0.5 and 1 M NaCl; 0.1 and 0.5 M Ca(NO3)2; 0.1 and 1 mM FeCl3; 10) has generally shown that: (1) the optimum pH is 4.5 for uranium and plutonium, (2) increasing salinity results in slightly
lower for uranium and significantly higher chemical recovery plutonium and (3) the presence of Ca(II) cations doesn’t significantly
affect the chemical recovery of both radionuclides. Contrary, the presence of Fe(III) cations ([Fe(III)] > 0.1 mM) results
in significantly lower chemical recovery for both radionuclides (<50%). The later is attributed to the formation of Fe(III)
colloids, which present increased chemical affinity for uranium and plutonium and hence compete with the radionuclide binding
by the resin. Nevertheless, the results indicate that the method could be successfully applied to a wide range of natural
Authors:R. Kumaresan, K. Venkatesan, R. Sajimol, M. Antony, T. Srinivasan, and P. Vasudeva Rao
Imidazolium nitrate anchored on poly(styrene-divinylbenzene) co-polymer, Im-NO3, has been synthesized and evaluated for plutonium purification. The results are compared with those obtained using Dowex
1 × 4 anion exchange resin. The distribution coefficient (Kd) of Pu(IV) increased with increase in concentration of nitric acid, reached a maximum at 8 M, followed by decrease in Kd values. Rapid ion exchange of Pu(IV) followed by the establishment of equilibrium occurred within 100 min of equilibration
and the data was fitted in to first order rate equation. Variation of distribution coefficient of Pu(IV) as a function of
exchange capacity and nitrate ion concentration suggest the involvement of anion exchange mechanism is responsible for extraction.
The apparent ion exchange capacity was 310 mg/g at 8 M nitric acid. The performance of the Im-NO3 under dynamic condition was assessed by column breakthrough experiments. Radiolytic degradation of Im-NO3 resin in presence and absence of nitric acid (8 M) was studied and the results are reported in this paper.
Authors:Lav Tandon, Kevin Kuhn, Diana Decker, Donivan Porterfield, Kenneth Laintz, Amy Wong, Michael Holland, and Dominic Peterson
Plutonium metal exchange programs operated by the Rocky Flats Plant were conducted from 1956–1989 to ensure quality and to
compare measurements in a plutonium metal matrix. Los Alamos National Laboratory (LANL) re-established the program in 2001
to assess the quality of analytical chemistry capabilities that support special nuclear material characterization. It is the
only program of its kind for the preparation and distribution of plutonium metal reference materials with a range of impurity
contents to multiple laboratories for destructive measurements of elemental concentration, isotopic abundance, and both metallic
and non-metallic impurity levels. This program provides independent verification of analytical measurement capabilities for
each of the seven currently participating laboratories, and allows any technical problems with analytical measurements to
be identified and corrected. This paper focuses on basic program elements and presents a summary of methods and results for
plutonium, uranium, neptunium, and americium, measurements.
Authors:Sachin Pathak, I. Pius, S. Mukerjee, Sangeeta Pal, and P. Tewari
Polyacrylhydroxamic acid resin synthesized by functionalization of polyacrylamide with hydroxylamine has been investigated
for the sorption of plutonium(IV) from carbonate medium, aiming at its application for the removal of plutonium from alkali
wash effluent generated during purification of TBP in PUREX process. Batch experiments have been carried out to determine
distribution coefficient of plutonium(IV) between this exchanger and various compositions of carbonate medium. Effect of the
concentration of sodium carbonate, sodium bicarbonate and pH of the solution on the distribution coefficient have been studied
to optimize the conditions for the uptake of Pu(IV) by this exchanger. Column experiments were carried out to determine the
practical capacity of the exchanger for plutonium. Elution studies were also carried out to recover the loaded plutonium from
the ion exchange column The exchanger displayed good exchange capacity for Pu(IV) from feed solution simulating the conditions
of carbonate wash effluent generated in PUREX process. The exchanger also exhibited fast elution of Pu, suggesting the feasibility
of using it for the recovery of Pu from carbonate based wash effluent.