Low-level radiochemical separations

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The initial fuel dissolution process may involve very corrosive materials and require expensive equipment. Americium, curium, and, under most conditions, neptunium, remain in the waste stream. Research and development studies suggest that it has promise for extracting americium and heavier actinides as well as residual amounts of uranium and plutonium. Ion exchange using both organic and inorganic exchangers finds wide application in radiochemical separations.

However, when organic exchangers are used, special attention must be paid to radiation effects on the organic materials, because none of them is totally resistant to radiation, and the degradation products affect the operation of the extraction system. From the hundreds of promising separation methods studied in the laboratory, PUREX has emerged over the years as the predominant method for commercial nuclear fuel reprocessing in a number of countries, notably France, the United Kingdom, Russia, and Japan.

The chemical basis for the PUREX process is that TBP selectively extracts uranium and plutonium when they are oxidized and in a complexant solution of high ionic strength, such as that provided by moderately concentrated nitric acid. Neptunium may also be extracted when conditions are adjusted properly. Plutonium is then selectively removed from the organic stream by contacting it with moderately concentrated nitric acid containing a reducing chemical in an operation called "stripping.

The decontamination factor DF is the ratio of the impurity concentration relative to the desired product before processing to that concentration after processing; in the case at hand, the ratio of fission product concentration in uranium or plutonium before processing to the same concentration after processing. Some plants employ an ion exchange step in place of the last plutonium extraction cycle, with similar results.

Other separation steps would have to be added to isolate any of the individual fission product elements in the waste stream. TRUEX exists in two versions, depending on the extractant type used. Both extractants are organophosphorus compounds dissolved in hydrocarbon solvents alkanes.

A third liquid phase 2 that is rich in the TRU elements may form when the TRU elements enter the alkane organic phase.

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To prevent this, TBP may be added to the organic phase. Third-phase formation must be avoided because its presence can cause nearly insuperable problems in the operation of continuous extraction equipment. It may sometimes be avoided by raising the temperature or by using aromatic or chlorinated hydrocarbon diluents instead of saturated hydrocarbons.

Isopropyl benzene has been employed as the solvent for this reason. However, solvents other than alkanes are often considered to have unacceptable flammability and health and safety problems. In many common solvent extraction processes, dilute nitric acid is used to back-extract or strip tetravalent actinide ions from the organic extractant into the aqueous phase.

Complexing agents that add more to the waste volumes to be treated are often used, and studies on how to avoid them are needed. Thus, there is a possibility that the system might be adapted to remove technetium from acidic wastes. It is not yet clear whether the CMPO extractant would be near-optimal for trivalent element separations.

Radioanalytical methods

The TRUEX process requires further research and development and production-scale testing before it can be considered to be fully available for use. In this process, amides, for example, dimethyldibutyltetradecylmethylamide DMDBTDMA 3 are dissolved in alkanes to extract TRU elements selectively, and with good separation factors, from nitric acid solutions.

Advantages claimed for this process are very low solubility in nitric acid, good extraction of metal ions without third-phase formation, and good thermal and radiolytic stability. An amide with improved properties is being sought for some of the more difficult separations, such as that required for Am III. A particular advantage of this class of solvents is that they can be incinerated completely, while, under ordinary circumstances, the ash inherent in the combustion of the phosphorous-based extractants is a significant waste product.

Process scientists in Russia and the former Czechoslovakia are investigating the use of cobalt dicarbollide as an extractant for cesium and strontium Mason et al. They have developed large-scale production methods for the reagent and have recovered megacuries of cesium and strontium from actual PUREX-processed acid HLW. The extractant possesses good radiation resistance, has good selectivity for cesium and strontium, and is stable over a wide range of acidities.

A potential drawback to the dicarbollide is that it must be dissolved in a polar solvent, which in turn has significant solubility in water. For applications to alkaline wastes as at Hanford tank wastes, engineering studies involving non-halogenated aromatic solvents will be required. Ion exchange processes have had considerable success in aqueous processing for removal of cesium and strontium from wastes. They are the basis of industrial processes for the separation of individual rare earth elements and other trivalent elements like the heavy actinides.

Organic ex. Third-phase formation can be a problem in the PUREX process also, but it is controlled with less difficulty by dilution. The reagents used for column regeneration and for elution of the desired components of the feed are considerable waste producers in many processes. Further discussion of these and other emerging separations technologies applicable to radioactive wastes is presented in Appendix D 4. Pyroprocessing is a major separations technology that has proved effective in recycling actinides in defense-related primarily plutonium-bearing materials and metallic fuel at EBR-II.

Plant designs based on preliminary research and development results indicate that pyroprocessing has the potential to reduce the size of plants and equipment needed Steunenberg et al. Other attributes of pyroprocessing plants such as their inherent high-temperature batch operations, when coupled with the preliminary nature of fundamental design assumptions, suggests that the indications of lower cost are very uncertain and not sufficiently reliable to form a basis for current decisions.

Another important advantage of pyroprocessing is its ability to handle fuels from the ATW and the ALMR that have been out of the reactor for much less than a year. The inventory build up is lower and the operations are on a smaller mass flow scale relative to PUREX. The nonaqueous nature of the process reduces the radiolysis problems. If specifically designed for such a purpose, pyrochemical processing of spent reactor fuel can provide high separation factors between members of chemically similar families of elements.

However, it requires structural materials that can withstand the high temperatures and the corrosive molten salts. A variety of processes that promise improved separations have been proposed on the basis of laboratory research. Only the major ones are mentioned here; more detail is provided in Appendix D.

The Integral Fast Reactor IFR fuel separation process, which had been under development at Argonne National Laboratory, is based on the selective electrorefining of uranium, plutonium, and heavier actinides from a molten cadmium solvent the anode into which they have been dissolved by anodic dissolution of spent IFR fuel.

The basis of electrorefining is that, under a given set of conditions, each metallic chemical element has a well-defined, nearly unique electrical potential at which it dissolves in a medium such as molten chloride salt, allowing elements or closely related element groups to be separated. Conversely, there is also a potential at which an ionic form of a metallic element is removed, or plated, from an ionic solution.

By adjusting electrical potentials appropriately, it is possible to transfer metallic elements such as uranium and transuranic actinides selectively from a metallic phase to an ionic phase and back again, and thus to effect their separation from one another and the bulk of the fission products. The IFR fuel separation process is designed to separate the transuranic actinides as a group and not produce an essentially pure plutonium stream. However, this also results in the plutonium product containing a significant amount of rare earths that reduce the worth of the fuel during recycle.

The electrorefining process somewhat resembles the salt transport process described in Appendix D , but the driving potential is provided electrically rather than chemically and can be controlled with greater precision. In the case of the IFR fuel separation process, actinide elements are dissolved electrochemically from spent fuel and deposited in a molten cadmium metal cathode. This increase is large enough to change the dissolution potentials so that uranium present in the cell either in the cadmium phase or in equilibrium with the ionic salt solvent is transported preferentially from the cadmium and deposited on a solid cathode.

On depletion of the uranium, the operating cell voltage rises slightly; plutonium with other actinide metals and some rare earth fission products is then transported across the ionic salt phase to the solid cathode. In the IFR process, the cathodes are changed at the onset of a cell voltage rise, effectively accomplishing separation of uranium from plutonium electromechanically. The more chemically active elements e. The more noble fission product metals e. The fission products remaining in the cell salt are discarded as waste.

Alternate technologies and processes which are in the initial stages e. Chang et al. All but one of the several potential separations processes proposed by the accelerator transmutation of waste ATW project would use a molten mixed-fluoride salt as a carrier for the primary target loop.

Introduction

The molten salt would be circulated through a side stream external to the target region, and separations would be performed on the molten salt to produce targets for reintroduction into the nuclear reaction zone. Separations processes for use in the ATW concept must be very tolerant of the intense radiation fields produced by the fission products present in the target fluid because of the proposed short cooling times and consequent high radiation fields.

It is doubtful that organic extractants could be used for these separations unless their process separation times could be reduced far below those achieved to date It has yet to be demonstrated that molten salt processes can achieve all the proposed separations to the degree desired. It was found that the other actinides and fission products could be held in the melt in sufficient concentrations for a successful reactor system.

For reprocessing it was found that some of the actinides and rare earths could be selectively removed from the melt by extraction with controlled concentrations of lithium metal in liquid bismuth followed by extraction with molten lithium chloride. The volatile fluorides were removed from the primary circuit molten salt with hydrogen fluoride and fluorine resulting in the removal and fractionation of iodine and uranium.

Technetium acts much like a noble metal and exists in a lower oxidation state that is highly insoluble in the molten salt. In the MSRE it was found that Pa could be extracted from the melt with 7 Li and bismuth in relatively pure form prior to decay to U. However, research and development studies on some of these processes for other applications showed that serious problems remained to be addressed before such a system would be viable Rosenthal et al.

Isolation of uranium from bulk impurities or fission products by volatilization has been demonstrated to be a practical approach that could be scaled to industrial levels. The separation possibilities range from the recovery of uranium from ore concentrates which is current practice to the recovery of uranium from a molten salt by using in-situ fluorination with elemental fluorine or perhaps the fluorinating agent O 2 F 2 to produce volatile uranium hexafluoride.

A number of other elements of interest, such as molybdenum, technetium, and iodine, also can be volatilized from spent fuel, introducing some complications into this isolation method. The individual elements may be separated by fractional distillation. This section considers the separations required for the major transmutation concepts that are discussed in detail in Chapter 4. Chemical processes are an integral part of any transmutation scheme to separate the radioactive components of the wastes into high purity fractions that can then be made into transmutation targets.

Such targets would be irradiated in a neutron flux having sufficient intensity and energy such that the radionuclides in the targets would be either transmuted or fissioned into stable elements or isotopes with substantially shorter half-lives at an acceptable rate. It is desirable, in some cases, that neutron irradiations take place at specific neutron energy and flux levels during relatively narrow time intervals in order to fission the isotope before it decays to a less easily destroyed nuclide. This is true in particular for those transmutation concepts that have high neutron fluxes, such as accelerators.

The narrow time window imposes severe requirements on the separations process. Figure shows the neutron capture paths of importance in the build-up of actinides, both for neutron fluxes commonly attained in reactors and for the very high neutron fluxes available from accelerators. The current amounts of actinides and fission products acceptable in the final LLW disposal matrix have been set rather arbitrarily by the proposers of various transmutation systems.

Radionuclide laboratory

For example, it is assumed that U. However, final LLW criteria for this purpose have not been established by the NRC and there is no generally accepted separations goal. Therefore, there can be no as-. The same comment holds for all separations processes associated with proposed transmutation systems. They are described below. Separation of plutonium and uranium from LWR spent nuclear fuel for future use in such reactors requires the partial separation of these elements from undesirable TRUs, fission products, and zirconium-alloy cladding hulls. With minor exceptions, all existing LWR fuel reprocessing plants have employed the PUREX process with the objective of producing pure plutonium and uranium.

There are no major reprocessing differences between spent fuels made with enriched uranium and those using recycled plutonium as the fissile isotope, the "mixed-oxide" MOX fuels.

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As already mentioned, the standard PUREX process does not separate americium and curium from the fission product wastes and separates neptunium only under special conditions. The modern PUREX process recovers a large enough fraction of uranium and plutonium to justify a commercially viable business and meet applicable environmental regulations about A separation methodology, either aqueous or pyroprocessing, could be used to extract plutonium and other individual actinides from spent LWR fuel for consumption in IFRs.

However, a great deal remains to be done in defining important parameters such as required actinide recoveries, processes for achieving them, and costs. The pyroprocessing approach that has been under investigation for this is promising; engineering-scale demonstrations of the head-end of the process have been started at Argonne National Laboratory-West in Idaho.


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This process, which employs direct electrochemical separation to purify the metallic reactor fuel from metal feed to metal product, could require fewer separations steps than aqueous processing when used on metallic fuels because the latter require that the metallic spent fuel be converted to an oxide or salt and then back again to metallic fuel in the course of recycling. An additional potential advantage is that it uses much smaller processing equipment because of the higher concentration of elements to be separated, although this is mitigated by the use of high temperatures, inert atmospheres, and batch processes.

Further discussion of the status and further development needs of pyrochemical processes are given in Appendix D pages Both aqueous and nonaqueous separations have been proposed for accelerator-driven reactor fuels. Such reactors could transmute fission products such as 99 Tc and I more rapidly than conventional reactors because they are projected to have higher thermal neutron fluxes see Figure One of the proposed ATW concepts features a target that is a slurry of TRU oxides in deuterium oxide heavy water, D 2 O that is processed in a continuously removed side stream of the circulating target stream.

This use of a fluid target and blanket material that can be processed continuously in a side stream is claimed as an advantage of the ATW system. However, there appear to be major hurdles to be surmounted in applying aqueous processes to the ATW system:. Two separate target loops would be needed, one to handle the uranium, plutonium, and higher actinide elements, and the other for the technetium and iodine. Criticality concerns mandate very low inventories in the circulating slurries and in the separation plant.

The ATW proposal involves first evaporating the deuterium oxide fuel slurry to dryness and then dissolving it in nitric acid to allow aqueous processing with organic extraction processes. The proposed use of continuous rather than batch processing may be overly optimistic due to the intense radioactivity: some interim storage is likely to be necessary to reduce the heat load resulting from the decay of short-lived radionuclides, since this heat could greatly complicate the aqueous separation processes. Radioactive decay power levels would have to be below about 10 watts per liter to prevent boiling of the aqueous solution; for instance, the heat load from any Cm produced watts per gram would require that no more than about 0.

These considerations may make the objectives of actinide fissioning and fission product destruction extremely difficult to obtain. It involves aqueous electrochemical oxidation of the carbon-coated fuel of their pebble bed reactor followed by a PUREX process. This cycle would benefit from the advanced processes that are being developed in connection with other proposed systems, as long as their fuel type can be efficiently processed and the active material dissolved in an aqueous solution.

No demonstration of this process system has yet been performed. A variety of nonaqueous separations, including fluoride volatility and molten salt processes, are also under study for. All pose challenging problems.

Radiochemical separation of mostly short-lived neutron activation products - INSPIRE-HEP

Previous research and development on fluoride-volatility and molten-salt processes revealed several serious difficulties in developing metal alloys that would resist corrosive attack by the tellurium fission product. New techniques are needed for operating separation cascades in molten-salt and molten-metal systems to achieve some of the very high separations needed. The separations research and development necessary for these accelerator-based transmutation schemes would be more extensive than that required by the IFRs because a wider variety of separations options must be considered.

A circulating molten salt that is processed by pyrochemical techniques may be the best solution for operating in the very high neutron flux and with the short-cooled materials in the ATW system, and it is this approach that has been receiving major emphasis at the Los Alamos National Laboratory.

No demonstration of this technology has yet been performed. Extensive laboratory study followed by demonstration of the technology would be required, as there has been little experimental research on the proposed ATW schemes. Combinations of aqueous and nonaqueous separation steps are also being considered by the research groups developing the ATW systems. There are special considerations at the interfaces between aqueous and pyrochemical steps involving the need to convert dissolved aqueous species to and from metals or salts, and these add to the costs of operations and waste generation.

Substantial improvements and innovations in separations technology suitable for very intense radiation fields are required to obtain the high degrees of separation and low process losses per cycle needed. Special attention would have to be given to losses in side streams and to waste streams from all the parts of the fuel cycle. More detailed findings and conclusions related to the separations aspects of the various approaches proposed are outlined in the paragraphs below. Conclusions and recommendations for defense wastes may be found in Chapter 5.

Improved separations technologies have the potential to reduce the amounts of high-level and TRU radioactive wastes by removing nonradioactive and low-activity bulk materials primarily uranium from spent LWR fuel waste. The usual PUREX process has overall losses of plutonium and uranium of a few tenths of a percent, yielding waste streams that are categorized as TRU waste. However, by using a modified PUREX process for enhanced plutonium, neptunium, and uranium separations and treating the resultant waste streams with advanced separations processes such as TRUEX, it might be possible to produce a waste with a to 1,fold reduction in those actinides.

These reductions are goals that can be reached with known but not fully developed technologies—at a price. In addition, selected long-lived e. Aqueous separations technologies based on new solvent systems and new ion exchange materials are in various stages of development. None has been demonstrated at the full engineering pilot-plant level.

For application to spent LWR fuel processing, an advanced technology e. Pyroprocessing involving electrorefining has been demonstrated on the several-kilogram scale with metallic IFR fuel. Kilogram-scale recovery has been performed with spent LWR fuel. Although research and development is not sufficiently advanced in this area to support firm conclusions, pyroprocessing of the type being developed for the IFR appears to be a satisfactory concept for processing LWR fuel if metallic fuels are to be used subsequently.

The ATW program goals for process losses per cycle are 0. It is assumed that those goals are sufficient to permit waste from the ATW to be treated as LLW not requiring disposal in a geologic repository. However, a major process development program would be required to meet them. The separations processes for the ATW are substantially more challenging than those under consideration for the other transmutation concepts. The difficulties arise from two factors. The first is the very. This time factor makes it extremely difficult to use organic reagents and aqueous systems for the major separation steps.

Excessive radiolytic degradation of the organic separations agents and water radiolysis would be unavoidable. Pyroprocesses based on molten-salt and molten-metal systems are more radiation resistant and would not be subject to the same problems. The second factor is the large number of separate streams that must be processed because actinides, technetium, and iodine must all be processed separately, and recovered essentially completely, for recycling. Because of the first factor, this is likely to be accomplished with less-familiar nonaqueous processes.

The ATW separations concepts that have been proposed are far short of being demonstrated processes; in fact, they are at such a preliminary stage of study primarily on paper that any judgment on their ultimate viability is necessarily premature at this time. Waste materials generated during processing and maintenance operations for the ATW would have to be treated for recovery to degrees never before achieved in large-scale systems in order to reach the low overall system losses required by the stated ATW goals.

Benedict, M. Pigford, and H. Nuclear Chemical Engineering , 2nd ed. New York: McGraw-Hill. Chang, Y. Christensen, D. Plutonium metal production and purification at Los Alamos. Carnall and G. Choppin, eds. ACS Symposium Series Washington, D. Coops, M. Knighton, and L. Pyrochemical processing of plutonium.

Culler, F. Bruce, I. Fletcher, H. Hyman, and J. Katz, eds. Hill, O. Scale-up problems in the plutonium separations program. By means of retaining the complex m Lu-DOTATATE in an apolar chromatographic column, the freed Lu can be separated by fluxing the column with a mobile phase with the proper polarity.

In this way, a new type of radionuclide generator is conceived, in which both the parent and the daughter nuclides are the same element. The generator will open new possibilities for the production and availability of the therapeutic radionuclide Lu and can bring significant growth in the research and development of Lu based pharmaceuticals. Reversed phase material, tC silica was purchased in the form of ready to use sep-pak cartridges Sep-Pak Plus tC18, usable for pH 2—8 , from Waters.

The completion of the reaction was checked using instant thin layer chromatography with acetonitrile: water as the mobile phase, and silica as the stationary phase. The column was manually filled with tC reversed phase silica waters. A slurry of tC silica in MeOH was added from one end of the column and the other end was connected with a vacuum pipe. The empty column has a volume of 0. The column was then equilibrated with the mobile phase for overnight before injecting the complex. The mobile phase and column were both temperature controlled to the desired temperature by a thermostatic circulation water bath Colora WK4 and a column water jacket Alltech.

Mobile phase flux of 0. The whole experimental setup was equilibrated for at least two hours with the mobile phase prior to loading of the complex. The complex was loaded on the manually filled tC column using a Rheodyne injector, with a mobile phase flow of 0. Once loaded, a column is used upto 3 months for doing the measurements. After a maximum period of 3 months, the column is flushed with pure methanol to remove all the loaded activity. All the fractions were measured on well-type HPGe detector. The efficiency calibration for different peaks was performed using a known activity of Lu source supplied by IDB Holland.

To minimize the error, all the vials were weighed before and after the fraction collection. In the continuous elution mode two flow rates were studied, 0. Each fraction was measured on the above mentioned well type germanium detector. The individual results and calculations can be found in supplementary information in sections S2 and S3. For flushing the accumulated activity the flow rate of 0. Detailed results and explanations are given in supplementary info S5. How to cite this article: Bhardwaj, R.

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Trends in Biotechnology 31 , —, doi: Blaauw, M. The holistic analysis of gamma-ray spectra in instrumental neutron activation analysis Download references. We are also grateful to Dr. Blaauw and to M. Lastly, the author would also like to acknowledge Ir. The manuscript was primarily written by R. Correspondence to P. This work is licensed under a Creative Commons Attribution 4.

By submitting a comment you agree to abide by our Terms and Community Guidelines. If you find something abusive or that does not comply with our terms or guidelines please flag it as inappropriate. Article metrics. Advanced search. Skip to main content. Subjects Medicinal chemistry Nuclear chemistry. Abstract Lu has sprung as a promising radionuclide for targeted therapy.

Introduction Lutetium Lu has emerged as a promising radionuclide for targeted therapy. Figure 1: Schematic representation of the decay process. Full size image. Figure 2: Separation of nuclear isomers Lu and m Lu. Results Continuous elution Initially, experiments with continuous flow of mobile phase or continuous elution were performed at different temperatures and mobile phase fluxes.