Extraction Of Plutonium Research Articles

A study of the extraction of plutonium from planchets by Aridus–ICP–SFMS

Two extraction processes of plutonium (Pu) on planchets from alpha spectrometry (AS) have been evaluated by inductively coupled plasma sector field mass spectrometry with a desolvator system (Aridus–ICP–SFMS). The samples were traced with known concentrations of 239Pu (1.2 × 103 fg) and 242Pu (2 × 103 fg) followed by an electrodeposition in planchets, according to the Hallstadius method. The processes of extraction were carried out with 50 mL of 0.36 mol L−1 HNO3 every 30 min up to 180 min in a glass beaker at 60 °C. The first process was on a hotplate and the second used an ultrasonic system. Finally, samples were evaporated to dryness, and resuspended in 10 mL of 0.72 mol L−1 HNO3 for evaluation. The results showed that at 120 min, a ~70 % recovery of 239Pu and a ~80 % recovery of 242Pu in both processes were obtained. The average recoveries of 239Pu and 242Pu at 180 min using the hotplate in plate were 93.4 ± 4.6 and 93.7 ± 4.2 % respectively, and with the ultrasonic system were 96.0 ± 4.3 and 98.2 ± 1.0 % respectively. In conclusion, both processes are suitable for Pu extraction, and Aridus–ICP–SFMS is an essential technique for the reassessments and quantification of Pu. In addition, procedural blanks spiked with 1 × 102 fg mL−1 U were prepared for each process, in order to study the contribution of the 238U on the background signal at m/z = 239, which was 0.5 ± 0.2 cps, indicating that the contribution of 238U on the 239Pu signal was negligible. Furthermore, this methodology can be applied to sample planchets with environmental, food, biological and nuclear origin, and thereby to avoid repetitive analysis when Pu concentration determined by AS are under minimum detectable activities.

Solvent extraction of Pu(IV) with TODGA in C6mimTf2N

Studies on the solvent extraction of Plutonium(Pu(IV)) from aqueous nitric acid by N,N,N′,N′-tetraoctyl-diglycolamide (TODGA) in 1-hexyl-3-methylimidazolium-bis(trifluoromethylsulfonyl) imide (C6mimTf2N) room temperature ionic liquid (RTIL) were carried out. It was found that Pu(IV) is extracted into RTIL phase as [Pu(NO3)(TODGA)]3+ through cation exchange mechanism. Extraction reaction equation is obtained by the influence of acidity and extractant concentration, and the parameters of thermodynamic equilibrium constant was calculated.

Flawed Nuclear Physics and Atomic Intelligence in the Campaign to deny Norwegian Heavy Water to Germany, 1942–1944

The military campaign to deny Norwegian heavy water to Germany in World War II did not diminish as the threat posed by heavy water in German hands dwindled, mainly because of excessive security among the Allies. Signs that Albert Speer (1905–1981) had decided in 1942 to stop the German atomic-bomb project were kept secret and ignored. Prominent Allied advisers like Leif Tronstad (1903–1945) and even Niels Bohr (1885–1962) were not told about the plutonium path to a German atomic bomb. Physicists did not brief advisers, decision makers, and Allied officers on how many years Werner Heisenberg (1901–1976) would need to accumulate enough heavy water (deuterium oxide, D2O) for an Uranmachine and then to extract and process plutonium for an atomic bomb. Had the flow of information been better, the military raids on the Norwegian heavy-water plant at Vemork could have been timed better, and the more costly of them could have been averted altogether.

Developing and evaluating di(2-ethylhexyl) orthophosphoric acid (HDEHP) based polymer ligand film (PLF) for plutonium extraction

A method was developed to produce a thin Polymer Ligand Film (PLF) using di(2-ethylhexyl) orthophosphoric acid (HDEHP) ligand for rapid extraction of plutonium. Ligands were incorporated into a polystyrene matrix to generate a smooth and durable surface for analyte extraction. PLFs were prepared with varying amount of ligands to find the optimimum ratio between the ligands and polymer structure to give best plutonium recovery. Also, the plutonium extraction is dependent on the concentration of nitric solution. This dependency was examined in the manuscript. The PLF samples were counted directly using alpha spectroscopy without any further chemical process to test the plutonium recovery. The alpha spectroscopy data shows that HDEHP based PLFs were effective in extracting plutonium from nitric acid solution. The surface characterization was performed with a scanning electron microscope (SEM) and digital autoradiography for defect examination of the PLF and plutonium distribution study, respectively.

Extraction of Americium(III), Plutonium(IV, V) and Neptunium(V) with Calixarenes

The distribution coefficients of Am iii , Pu v , Pu iv and Np v on extraction from aqueous media with solutions of calix [4] arenes in 1,2-dichloro-ethane and their dependence on the structure of the macrocycles have been determined.

Studies on solubility of TBP in aqueous solutions of fuel reprocessing

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.

Lessons Learned from the North Korean Nuclear Crises

How Did North Korea Develop the Bomb?The promise and peril of nuclear energy share a common technological foundation. Pursuit of a civilian fuel cycle-making fuel, building reactors to burn the fuel, and maintaining the back-end to deal with nuclear waste, including the option of extracting some of the valuable by-products from burning reactor fuel-enables nations to develop the capability to make bomb fuel, either highly enriched uranium (HEU) or plutonium. North Korea mastered the plutonium fuel cycle ostensibly for nuclear power and then used it to build the bomb.This brief review of North Korea's acquisition of nuclear capabilities will only touch on the important political milestones that helped to shape it; a more complete discussion will be presented in the next section. Kim Il-sung, the country's founding father, laid the foundation for nuclear technology development in the early 1950s. The Soviet "Atoms for Peace" initiative, modeled after President Eisenhower's initiative of the same name, enabled several hundred North Korean students and researchers to be educated and trained in Soviet universities and nuclear research centers. The Soviets built a research re-actor, the ITR-2000, and associated nuclear facilities at Yongbyon in the 1960s. North Korean specialists trained at these facilities and by the 1970s were prepared to launch a nuclear program without external assistance.North Korea's decision to build gas-cooled, graphite-moderated reactors was a logical choice at the time for an indigenous North Korean energy program because gas-graphite reactors can operate with natural uranium fuel and, hence, do not require enrichment of uranium.3 Although North Korea may have experimented with enrichment technologies, commercial enrichment capabilities were beyond its reach and difficult to acquire.4 North Korea's ambitious program began with an experimental 5 megawatt-electric (MWe) reactor, which became operational in 1986. Construction of that reactor was followed by a scaled-up 50 MWe reactor and a 200 MWe power reactor, although neither was ever completed.North Korea quickly mastered all aspects of the gas-graphite reactor fuel cycle. It built fuel fabrication facilities and a large-scale reprocessing facility, which enabled extraction of plutonium from spent fuel.5 Unlike the Soviet built research facilities, the new facilities were built and operated without being declared to or inspected by the International Atomic Energy Agency (IAEA). Pyongyang had no legal obligation to declare these facilities because it was not a member of the Nuclear Non-Proliferation Treaty (NPT). American reconnaissance satellites picked up signs of the reactor construction in the early 1980s and the reprocessing facility in the late 1980s. It was not until 1989, when South Korea leaked American satellite data of the reprocessing facility that the international community first became aware of and concerned about North Korea's indigenous nuclear program. The concern stems from the fact that gas-graphite reactors are capable of producing weapons-grade plutonium while generating electrical power and heat. So, whereas Pyongyang's choice of gas-graphite reactors for its energy program was logical, it was also the best choice to develop a nuclear weapons option.In parallel, North Korea asked the Soviets to build light water reactors (LWRs) to help meet North Korea's energy demands. North Korea joined the NPT in 1985 because the Soviets made consideration of LWRs contingent upon joining the Treaty. These reactors, though, never materialized because of the demise of the Soviet Union. Pyongyang kept inspectors out of its new facilities until 1992, by which time it had all of the pieces in place for the plutonium fuel cycle. This move coincided with several diplomatic initiatives and President George H.W. Bush's decision to withdraw all American nuclear weapons from South Korea. By this time, the 5 MWe experimental reactor produced electricity and heat for the local town, as well as approximately 6 kilograms (roughly one bomb's worth) of weapons-grade plutonium per year. …

Recovery of Uranium, Plutonium, and Neptunium from High-Level Waste Solutions prior to Actinide Partitioning

AbstractThe prior removal of uranium along with plutonium and neptunium from high-level waste (HLW) can simplify the subsequent actinide partitioning steps. In this context, extraction and stripping behavior of uranium, plutonium, and neptunium was investigated employing 1.1-M solutions of tri-n-butyl phosphate (TBP) and N,N-dihexyl octanamide (DHOA) dissolved in n-dodecane as extractants in the presence of 6 g/L U relevant to pressurized heavy-water reactor (PHWR) high-level waste (HLW) conditions. Reductants such as acetohydroxamic acid (AHA) and hydroxyurea (HU) were evaluated for selective stripping of plutonium and neptunium over uranium from the loaded organic phases. These studies indicated the possibility of quantitative recovery of U(VI), Pu(IV), and Np(IV) from HLW solution employing either DHOA or TBP as extractant. Integration of this step in the minor actinide partitioning simplifies the treatment of HLW with the specially designed extractants like octyl(phenyl)- N, N-diisobutyl carbamoyl me...

Extraction Experiments after Radiolysis of a Proposed GANEX Solvent-The Effect of Time

In a solvent intended for use within group actinide extraction CyMe4-BTBP and TBP are combined together in cyclohexanone. After irradiating the solvent in acid contact a latency period before extraction reduced the extraction capability of plutonium. This reduction was larger when the solvent was kept in contact with the acid after irradiation. The decrease in plutonium extraction could be an effect of both a reduced extraction performance of CyMe4-BTBP caused by hydrolysis as well as a shift in oxidation state of the plutonium as it was shown that oxidizing Pu(IV) to Pu(VI) decreased the extraction by the solvent.

Recovery of Uranium and Plutonium from Waste Matrices Using Supercritical Fluid Extraction

Supercritical fluid extraction (SFE) of plutonium in its nitrate form from actual waste, i.e. plutonium bearing cellulose matrix was demonstrated using 0.1 litre capacity extraction vessel. Complete recovery of plutonium was demonstrated using modified supercritical carbon dioxide (Sc-CO2), i.e. Sc-CO2 containing octylphenyl-N, N-diisobutyl-carbamoyl- methylphosphine oxide (φCMPO). Near complete recovery of uranium was demonstrated from simulated waste matri- ces, i.e. uranium bearing teflon, glass and cellulose matrices using preparative scale SFE, i.e. from 1 litre capacity ex- traction vessel. The recovery of uranium was established using Sc-CO2 modified with acetyl acetone.

Possible Options for Uranium-carbide SNF Processing

The dissolution of uranium carbide pods as surrogate of UC fuel was performed by adding calcium metal in nitric acid, individually or in conjunction with oxide WWER's fuel. Extraction tests realized with the solution coming from the dissolution of the uranium carbide pods under PUREX conditions demonstrated that the plutonium remained in the aqueous phase. According to literature, it was found that it is connected with the formation of water-soluble organic compounds during the disslution. Extraction test performed on solutions coming from the dissolution a mix of uranium-carbide and oxide fuel (WWER) showed a complete extraction of uranium and plutonium and no plutonium retention of in the organic phase at the stripping step.

Validation of the flow-sheet proposed for reprocessing of AHWR spent fuel: counter-current studies using TBP

ABSTRACT A solvent extraction based flow-sheet, for reprocessing of spent fuel arising from Advanced Heavy Water Reactor (AHWR) for separation of uranium, plutonium and thorium using 5% TBP in n-dodecane, has been tested using laboratory scale mixer-settlers. Simulated feed solution containing thorium, uranium and plutonium in the concentration levels expected in feed solutions of AHWR spent fuel reprocessing is prepared using plutonium, natural uranium and thorium nitrate. Quantitative extraction of uranium and plutonium is achieved under experimental conditions leaving bulk of thorium in the raffinate. Co-extracted thorium from the organic phase is scrubbed using 3.00 M HNO3. Separation of plutonium from uranium is achieved by chemical reduction employing a mixture of hydroxylamine nitrate and hydrazine nitrate in nitric acid. Results show quantitative partitioning. Uranium from the plutonium lean organic phase is stripped using 0.01 M HNO3. The results clearly established the validity of the proposed f...

Evaluation ofN,N-dihexyloctanamide as an alternative extractant for spent fuel reprocessing: batch and mixer settler studies

ABSTRACT Solvent extraction studies carried out at BARC, India on the evaluation of N,N-dialkyl amides as alternative extractants to tri-n-butyl phosphate (TBP) for reprocessing of spent fuel have suggested that straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels. This paper presents the batch as well as mixer settler studies for uranium and plutonium extraction/stripping to evaluate DHOA vis a-vis TBP for the reprocessing of Pu rich fuels. These studies showed that uranium extraction using DHOA as extractant was comparable to that of TBP; however, it displayed better stripping behavior than TBP. Plutonium extraction behavior was better in the case of DHOA as compared to that of TBP. However, Pu stripping data indicated towards the need of reducing agent in the stripping cycle for both the extractants.

Purification of americium from assorted analytical waste in hydrochloric acid medium

During the simultaneous extraction of plutonium and uranium using anion exchange chromatographic technique from analytical waste in hydrochloric acid medium, 241Am which is invariably present in the plutonium bearing fuel samples remains in the effluent. A two step separation scheme was developed for the recovery and purification of Am from the assorted waste to facilitate the disposal of large volume of aqueous waste and the purified Am solution was utilized for spectroscopic investigation. The separation scheme involved solvent extraction using 0.1 M TODGA + 0.5 M DHOA for separation of americium from Fe, Pb, Ni and Na followed by extraction chromatographic technique using CMPO on inert support as stationary phase for separation of Ca from Am. A systematic study on the extraction behavior of Am from hydrochloric acid medium revealed that out of four extraction systems well known for actinide partitioning namely 0.1 M TODGA + 0.5 M DHOA, 1 M DMDBTDMA, 0.2 M CMPO + 1.2 M TBP and 30% TRPO, only 0.1 M TODGA + 0.5 M DHOA extracts americium from 7.5 M HCl feed acidity. A comparative study involving CMPO solvent extraction and column chromatographic technique revealed that elution of Am from column is satisfactory as compared to inefficient stripping of Am from organic phase in solvent extraction technique using 0.1 M HNO3. The purity of the final solution was checked for 17 elements of interest and was found to be 98% pure, while the overall recovery of this two step separation scheme was found to be 95%.

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

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.

Solid-phase extraction of plutonium(IV) an americium(III) using N-benzoylphenylhydroxylamine and its derivatives

The recovery and separation of plutonium(IV) and americium(III) by solid-phase extraction (SPE) on alkylated silica gel S16 modified with N-benzoylphenylhydroxylamine (BPHA) and with its derivatives was studied. BPHA was modified by introducing into the p-position of the phenyl ring of electronactive substituents that differ in their hydrophobicity: CH3, Ph, Cl, F, and NO2. The SPE of plutonium(IV) and americium(III) was studied in the range of acidities from pH 8 to 1 M HNO3. The recovery and separation of these elements was shown to depend on the nature of the substituent, aqueous acidity, and the preparation of S16 to SPE experiments.

Reagent-based remediation of soil contaminated with transuranium elements

The results of studies on selecting remediation solutions for removing plutonium and americium from soil are presented. The effect of the type and concentration of different reagents on the degree of leaching of radionuclides is studied. It is shown that acid solutions extract americium from soils. Reducing media must be used to extract plutonium from soils efficiently. The conditions for obtaining 90% reagent-based remediation of soils are determined.

Solid-phase extraction of plutonium in various oxidation states from simulated groundwater using N-benzoylphenylhydroxylamine.

Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98–99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90–95% and 75–80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45–50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (K d) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80–90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents.

Rapid separation and extraction of radioactive analytes onto filters and surfaces

Abstract This paper presents a radiochemical procedure that combines radiochemical sample preparation and analyte separation to provide a rapid sample preparation method. Extractive ligands are prepared in a polymer matrix. This system can be immobilized on a surface or on a filter. Filters allow rapid sample uptake due to the movement of the analyte through the filter material. These approaches offer fast uptake kinetics, high material loading, and good selectivity. These systems have been demonstrated for the rapid extraction of plutonium and americium from solution and their subsequent analysis by alpha spectrometry, with no additional sample preparation required, thus eliminating extensive wet chemical processes that would otherwise be necessary. We have evaluated the extraction efficiency that can be obtained and have shown that these systems can provide for alpha spectrometry peak widths near what can be obtained using traditional techniques such as electroplating. We have also demonstrated Polymer Ligand Films (PLFs) for the selective extraction of plutonium from a solution prepared from a soil digest.

Perfluorinated phosphine oxide and sulfides as extractants for heavy metals and radionuclides

New highly fluorinated monodentate and bidentate phosphine oxide compounds of the type {CF(3)(CF(2))(n)CH(2)CH(2)}(3)PO (n = 5, 9) and [{CF(3)(CF(2))(5)CH(2)CH(2)}(2)P(O)CH(2)CH(2)P(O){CH(2)CH(2)(CF(2))(5)CF(3)}] have been prepared. Their ability to extract a number of metals and radionuclides from aqueous solutions into perfluorinated solvents has been established and the extractable species investigated. All extractants extract the metals As(V), Cd(II), Co(II), Cr(VI), Hg(II), Pb(II), and Sn(II) with >75% removal. In addition, the radioisotopes (90)Sr(II), (133)Ba(II), and U(VI) have been investigated, whilst (59)Fe(III) has been used to model the extraction of plutonium. (133)Ba(II) shows a high distribution ratio for monodentate phosphine oxides, whilst for UO(2)(2+) and (59)Fe(III) bidentate phosphine oxides are superior.