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Macroreticular anion exchange resin cleanup of tbp solvents.

AboutThis article is published in Transactions of the American Nuclear Society.The article was published on 1972-01-01 and is currently open access. It has received 6 citation(s) till now. The article focuses on the topic(s): Ion exchange & Ion-exchange resin.

Topics: Ion exchange (56%), Ion-exchange resin (53%), Fission products (51%), PUREX (51%)

Summary (3 min read)

DISCLAIMER

  • Portions of this document may be illegible in electronic image products.
  • Images are produced from the best available original document.
  • Typical applications include purification of aqueous uranium feedstocks in uranium refineries, recovery of [1] uranium and/or plutonium from metallurgical scrap [2, 3] and, most importantly, the Purex process for reprocessing all kinds of .irradiated nuclear reactor fuels. [4].
  • Chemical and radiolytic degradation of TBP extractants and solvent treatment procedures have been reviewed by several authors. [5-8].
  • Progress and findings of these latest resin solvent treatment studies are highlighted'in this paper.

Kinetics of sorption of fission products and HDBP

  • From used Purex solvent by A-26 resin were significantly.
  • Kinetics of fission product and HDBP uptake by A-26 resin also increased with decreasing resin particle size.
  • HF solutions were best for eluting fission products and HDBP from A-26 resin.
  • High capacity of A-26 resin for sorbing extractant impurities was indicated in very preliminary column runs.
  • Physical and chemical properties of the effluent solvent in these runs were equal to or superior to those of Hanford Purex plant carbonate-washed material.

MATERIALS

  • As-received Cl-form A-26 resin was converted to the OH-form by treatment with excess 4M NaOH, washed with water, and dried in air.
  • Prior to use with 1CW solution, all resin beds were classified by upflow of water.
  • Both upflow and downflow conditiohs were used with NaOH eluents.

PUREX PROCESS SOLVENTS Flow Rate Tests

  • Table II summarizes conditions and results of column runs made to study effects of flow rate and solution residence time on A-26 resin cleanup of Purex process solvent.
  • With one exception, these runs were made at a bed height:diameter ratio of 4.
  • Such solvent was produced only after breakthrough of fission product activity.
  • In plant-scale operation, eluates resulting from regeneration of A-26 resin beds would become, after evaporation, part of the Purex process high-level liquid waste stream.
  • Calculations to establish economically permissible frequency and volume have not been made, however.

Miscellaneous Observations and Tests

  • The authors have emphasized application of OH-form Amberlyst A-26 resin for cleaning up used Purex process solvent.
  • But other macroreticular strong base anion exchange resins (e.g., Amberlyst A-29, J. T. Baker Company A-641, etc.) can also be used for this purpose.
  • In the as-received Cl-form A-26 resin does not efficiently sorb fission products from Purex 1CW solvent (Table VI ).
  • The yellow color bodies are thought to be nitration products of the NPH diluent, but they have not been positively identified as such.
  • Sorption of these yellow colored compounds by A-26 resin is in line with manufacturer's claims for this resin. [10].

Limited test data shown in

  • Straight chain paraffin mixtures (e.g., NPH) suitable for use as a diluent for TBP have been commercially available only since about 1966.
  • Before then, commonly used diluents (e.g., Soltrol 170, Shell E-2342, etc.) contained various amounts of branched paraffins, olefins, and napthenes.
  • Abundant experimental evidence exists [5- to account for the removal of the zirconium and niobium.
  • Conversely, the authors find in both batch and column tests that Distribution ratios for sorption of I2 from 1CW and laboratoryprepared 30% TBP-NPH solvents were only 2.8 and 12.2, respectively.
  • Much of the iodine which may be present in TBP extractants used with highexposure fuels could be there as organic iodides.

Resin Application Schemes

  • Depending on economic factors and operating philosophies, there may be as many ways to use A-26 resin in cleaning up degraded Purex process solvents as there are Purex plants.
  • One solvent wash step and its attendant high-level aqueous waste are thus eliminated.
  • Scheme 2 features resin treatment of both first and second cycle solvents.
  • Counterbalancing this advantage, however, is the likely need for frequent resin regeneration and/or replacement of the resin used to treat first cycle solvent.
  • Detailed economic calculations (which have not been made) are required to decide the merit, if any, of Scheme 2.

Application of macroreticular resins in tailend cleanup

  • Of carbonate-washed first cycle solvent is, potentially, the most efficient way to take advantage of their favorable properties.
  • Such washing will remove the bulk of the fission product activity, essentially all .the HDBP and any entrained HN03, and provide an effective buffer zone to negate effects of periodic coextraction cycle upsets.
  • The overall effect should be to ensure a very long useful life for the A-26 resin bed before its replacement becomes necessary.
  • Operability of the "one solvent" system, particularly the use of resin-treated .first cycle solvent in the second uranium cycle, has not been demonstrated.
  • Loaded resin can be incinerated (see p. 33) or destroyed by reaction with HN03-H2S04 solutions according to the process being developed by Westinghouse Hanford Company workers. [21].

PLUTONIUM RECLAMATION FACILITY SOLVENT Conceptual Resin Solvent-Treatment Schemes

  • A solvent extraction process involving extensive product reflux is used in Hanford's PRF to recover and purify plutonium from various types of metallurgical scrap.
  • Sufficient resin capacity would be provided in both cases to treat used 8 solvent for one week before resin replacement.
  • Spent resin would be incinerated at about 760 °C in the present PRF in- [22] cinerator, and plutonium would be recovered from the ash by standard HN03-HF dissolution techniques.
  • Also, carbonate washing provides a buffer zone to ensure removal of HN03 and U02(N03)2, both of which A-26 resin also sorbs considerable yellow color from both CUW and CXW streams.
  • Exotherms occur at about 325 and 500 °C in the differential thermal analysis curve for OHform A-26 resin ; at about 550 °C the resin burns completely to leave a small amount of white ash.

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Abstract: A hydrazine oxalate [(N2H5)2C2O4] solution was used as an alternative wash method for cleaning 30% tributyl phosphate–70% normal paraffin hydrocarbon solvent. Experimental evidence shows the (N2H5)...

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Abstract: The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from saltmore » waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables.« less

18 citations


Journal ArticleDOI
Abstract: Several secondary cleanup procedures have been tested for hydrolytically and radiolytically degraded TRUEX process solvent (0.2 M. n-octyl(phenyl)N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO)-1.2 M tributylphosphate (TBP) in n-dodecane). Sodium carbonate scrub was used as primary cleanup. For the secondary cleanup macroporous anion exchange resins and other solid adsorbents, such as goethite (a-FeOOH), alumina and activated charcoal were used. The effectiveness of a cleanup procedure was established by its capability to restore the original americium(III) distribution ratio from low HNO3 concentration, that is characteristic of pristine process solvent. Further information was obtained from the measurement of up to seven successive AmfJIT) distribution ratios with the regenerated solvent, using the stripping conditions of the TRUEX process. Although all the procedures tested proved to be effective in removing most of the unwanted acidic products from the degraded solvent, the use of a stron...

18 citations


Journal ArticleDOI
Abstract: Solid sorbents, alumina, silica gel, and Amberlyst A-26 have been tested for the cleanup of degraded TRUEX-NPH solvent A sodium carbonate scrub alone does not completely remove acidic degradation products from highly degraded solvent and cannot restore the stripping performance of the solvent By following the carbonate scrub with either neutral alumina or Amberlyst A-26 anion exchange resin, the performance of the TRUEX-NPH is substantially restored The degraded TRUEX-NPH was characterized before and after treatment by supercritical fluid chromatography Its performance was evaluated by americium distribution ratios, phase-separation times, and lauric acid distribution coefficients 17 refs, 2 figs, 5 tabs

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Reference EntryDOI
04 Dec 2000
Abstract: The chemical reprocessing of discharged nuclear reactor fuel allows the large quantities of energy found in uranium and plutonium to be recycled back to the reactors while separating out the fission products as a waste stream. The PUREX process, used to recycle nuclear reactor fuel, was developed as part of the U.S. defense program. This process is described. The world's principal fuel reprocessing centers are located in the U.K., France, and Japan. Smaller units are operating in India. No chemical reprocessing takes place in the United States. Chemical reprocessing involves fuel decladding, mechanically cutting or shearing the tubing in which the nuclear fuel is encapsulated, and fuel dissolution, prior to chemical separation of the nonradioactive materials, the fission products, and the unburned fuel. Several types of countercurrent solvent extraction equipment are used to separate the useful components. The uranium and plutonium products are converted to oxides from which recycle fuel is fabricated. The wastes are safely stored for future disposal. Keywords: Separation processing; Nuclear reactors; Reprocessing strategy; Fuel characteristics; Product conversion; Waste handling; Fuel shear; Liquid waste storage tanks; Liquid-liquid contactors

3 citations