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Macroreticular ion exchange resin cleanup of purex process tbp solvent.

W.W. Schulz
01 Jan 1970-
About: The article was published on 1970-01-01 and is currently open access. It has received 4 citations till now. The article focuses on the topics: Ion exchange & PUREX.

Summary (2 min read)

Jump to: [DISCLAIMER] – [Over 240 bed volumes of• unwashed] – [MATERIALS] – [DISTRIBUTION RATIO TESTS] – [SOLVENT QUALITY TESTS] – [Elution Tests] and [Flow Rate Effects]

DISCLAIMER

  • Portions of this document may be illegible in electronic image products.
  • Images are produced from the best available original document.
  • The capacity of macroreticuZar resins for adsorbing ---..-Il .*.. extractant impurities is very high judging from batch and co Zumn data.

Over 240 bed volumes of• unwashed

  • Hanford Purex pZant first cycZe soZvent were passed downftow (at 40 C and 4 bed voZumes/hr) through a 59-ml bed,of 14 to 50 mesh res€n without any detectabZe breakthrough of impurities.
  • AZZ the effZuent soZvent was water-white as opposed to the faint-yezzow cozor .of the feed; its fission product content and pZutonium retention number were both substant€aZZy Zower than typical washed (atkaZine permanganate) pZant soZuent.
  • A major disadvantage of these methods, however, is that they generate large volumes of radioactive aqueous waste which must be stored or otherwise treated as high-level waste.
  • These large pores do not disappear when the swelling solvent, 2 ARH-SA-58 RD water, is removed; hence, macroreticular resins are especially suited for use with non-aqueous, even non-polar, solutions.
  • Changes in solvent quality and stability significantly enhance applicability of ion exchange solvent cleanup procedures, also known as These favorable.

MATERIALS

  • Rohm and Haas Company macroreticular resins were used throughout.
  • Amberlyst-15 (cation exchange) and Amberlyst A-21 (weak base anion exchange) resins were used in the as-received (H+-and OH--forms, respectively) condition.
  • Amberlyst A-26 and A-29 (strong base anion exchange) resins were converted from the as-received chloride-to the hydroxideform by exhaustive washing with 42 NaOH.
  • Air-dried resins were screened (U.S. Standard Sieve series) to obtain fractions encompassing desired particle sizes for use in batch equilibration tests.

DISTRIBUTION RATIO TESTS

  • Two-gram portions of screened, air-dried resin were contacted (30 min, 25 C; mechanical stirring) twice with fresh 10-ml portions of 30 percent TBP-NPH and then (at various times and temperatures) with 10 ml of either plant 1CW or laboratoryprepared 30.percent TBP-0.054M DBP-NPH.
  • Initial and final liquid phases from the last contact were analyzed either for fission product content [gamma energy pulse height analyses with NaI or Ge(Li) detectors] or for DBP [Beckman Automatic Titrator; derivative mode titration with alcoholic KOH].
  • Distribution ratios (Kd) for the loading step were calculated as ARH-SA-58 RD amount of material on resin per gram of air-dried resin Kd -amount of material in solution per milliliter of solution'.
  • After classification by upflow of water, the bed height was about 21 cm corresponding to a bed volume of 58 ml.

SOLVENT QUALITY TESTS

  • Characteristics of the strong adsorption of HDBP by strong base macroreticular anion exchange resins are shown in Figure 5 , page 19.
  • Kinetic effects operative in this case and their variation with temperature and particle size are similar to those noted for uptake of fission products.
  • In the Japanese work cited earlier, in addition to HDBP various other acidic components (e.g., monobutylphosphoric acid and carboxylic acids) present in degraded TBP solutions also reported strongly to the resin phase.
  • Indirect evidence for adsorption of acidic ARH-SA-58 RD components other than HDBP and fission products from Purex 1CW solution was obtained in column runs; this evidence is considered later.

Elution Tests

  • Various reagents were screened on a batch basis to determine their ability to elute fission product activity from loaded macroreticular resin.
  • Conversely, no completely satisfactory elutriant for removing 106 Ru-106Rh activity from the loaded macroreticular resin has yet been found.
  • Of the reagents tested for this purpose, NaOH and HN03-HF solutions appear best and were used in column tests.

Flow Rate Effects

  • Illustrative of the kinetic aspects of the new solvent treatment process is the way fission product absorption increases with decreasing feed flow rate (Figures 6 and 7 , pages 20 and 21).
  • Throughout both loading cycles, the fission product content of the organic effluent remained approximately constant.
  • The Pu retention number like the analogous "Z" and "H" (1 3) numbers has traditionally been considered a sensitive measure of the presence of deleterious diluent and/or TBP " degradation products in used Purex process solvent.
  • Also, quality of the product obtained by the ion exchange procedure is equal or superior to that of solvent washed with conventional alkaline permanganate solutions.
  • Improved techniques for eluting radioruthenium from the loaded resin are also needed.

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RECEIVED .BY DTIE JUL 31
1970
DOCUMENT
NO·
ARH-SA-58 RD
UNCLASSIFIED
DATE August 1,
1970
COPY NO.
C L A S S I F I C A T I O N •
ISSUIN.G FILE
Atlantic
Richfield
Hanford
Company
i
RICHLAND, WASHINGTON
TITLE AND
AUTHOR
CONf-1104/9-- L
MACRORETICULAR ION EXCHANGE RESIN
CLEANUP
-.2
OF PUREX PROCESS TBP SOLVENT
'-I..
By - ,=,.+ S
6 "t B
Wallace
W.
Schulz
i. CO m i..
DISTRIBUTION
NAME
6
t
BUILDING
AREA
A. E. Barber
(3)4.1-
71-U 200-West ARHCO
M. H.
Campbell
234-5
200-West ARHCO
W. W.
Schulz 222-S 200-West
ARHCO
R. E.
Tomlinson
2704-E
200-East
ARHCO
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UNCLASSIFIED
ARH-SA-58 RD
/ p
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MACRORETICULAR ION EXCHANGE RESIN CLEANUP
OF PUREX PROCESS TBP SOLVENT
BY
Wallace W. Schulz
Separations Chemistry Laboratory
Research and Development
Chemical Processing Division
August 1, 1970
ATLANTIC RICHFIELD HANFORD COMPANY
RICHLAND, WASHINGTON
THIS
DOCUMENT
CONFIRMED AS
UNCLASSIFIED
DIVISION OF CLASSIFIC4TIO
a N. lutjjab*K#dib/J--
BOYATE -
To be presented at .
International Solvent Extraction Conference, 1971
The Hague, Holland
April 19, 1971
DISTRIBUTION OF
THIS
DOCUMENT IS UNT.TMITED
Operated for the Atomic Energy Commission by
Atlantic Richfield Hanford Company under Contract #AT(45-1)-2130
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Citations
More filters
Book Chapter•DOI•
Aqueous Reprocessing by THOREX Process

[...]

P. V. Achuthan1, A. Ramanujam•
Bhabha Atomic Research Centre1
01 Jan 2013
TL;DR: In this paper, the major steps involved in this generic process and the several possible variations in the process flow sheet that can be utilized to meet the different end objectives are discussed, along with the practices followed at different facilities to treat bulk quantities of aged 233U and their recycle during fuel fabrication.
Abstract: In nature, thorium exists mainly in a monoisotopic form with no fissile isotope for its use as fuel in nuclear reactors. Though a fissile isotope of uranium, 233U is formed by neutron irradiation of fertile thorium in reactors; the subsequent reprocessing of irradiated thorium for recovery, purification, and further handling of 233U product (accompanied by 232U) has remained a challenge because of the complex radiological problems associated with irradiated thorium. During irradiation in the reactor, 233Pa with a 27 day half-life is formed by (n, γ) reaction of 232Th. Its complete decay to 233U is to be ensured prior to reprocessing for maximum recovery. The (n, 2n) reactions encountered during the irradiation of Th give rise to long-lived 231Pa and rather short-lived 232U (~70 years). The 232U and its hard beta gamma emitting short-lived daughters in the separated 233U and the 229Th and 228Th in thorium contribute to the radiation dose of these products. TBP has been the most widely used extractant in the nuclear industry for hydro metallurgical and reprocessing applications and hence, TBP-based THOREX process, in its various forms has become the natural choice to treat irradiated thorium in plants meant for PUREX process, on a campaign basis. Different flow sheets have been used and fine-tuned to meet the specific processing requirements of irradiated Th from different reactor systems based on the type of thoria target/fuel and the cladding under treatment, their irradiation, and cooling history and the end objectives, namely, separation and purification of Th/233U/233Pa in short cooled fuels, or Th and 233U in long cooled fuels/targets or 233U alone from the fuel target matrix and the final product decontamination factor aimed at. Thus the process has evolved over the years from the experience of individual plants that had conducted these pioneering earlier campaigns. This chapter details the major steps involved in this generic process and the several possible variations in the process flow sheet that can be utilized to meet the different end objectives. The different head end treatment options, fluoride ion catalyzed nitric acid dissolution of the fuel, the different TBP-based cycles for the extraction, individual separation and purification of Th/233U, extraction behavior of Th/233U/233Pa and fission products during this step, the solvent management in the cycles, third phase formation tendencies of Th with TBP, the solvent degradation and its consequences in the process are some of the topics covered. The global status on thorium utilization along with country-wise practices and experiences on thorium reprocessing, and recent developments in this domain have been briefly reviewed. Just prior to 233U product reconversion and fuel fabrication, as of now, a chemical separation of the longer lived daughters of 232U (like 228Th and 224Ra) is required to control the personnel exposure during these steps. The activities of short-lived, but strong gamma emitters in the chain, 208Tl and 212Bi, depend on 228Th, the nuclide with longest half-life in the chain. Techniques based on ion exchange, solvent extraction, or precipitation routes are followed for purification of 233U product from 228Th and for removing the residual thorium accompanying product. These are summarized along with the practices followed at different facilities to treat bulk quantities of aged 233U and their recycle during fuel fabrication. Third-phase formation encountered in TBP extraction of Th and Pu has been discussed in detail as it has serious repercussions on process performance and safety. In this context, many homolog neutral organophosphorus esters in the TBP family and alkyl amides have been tested as substitutes for TBP and some of them are reported to have better performance in terms of Th/U separation factors and higher thorium loading. Their status is presented. Major areas that need further developmental efforts for the process to succeed on industrial scale have been identified.

5 citations

Journal Article•DOI•
Comparison of the performance of solvent wash reagents used for the primary cleanup of degraded PUREX solvent

[...]

Satyabrata Mishra, C. S. Mukhopadhyaya, Chayan Patra, Vishnu Anand Panantharayil, Puspalata Rajesh, Dasari Sivakumar, Narasimhan Desigan, K. Dhamodharan, P. Velavendan, R. Rajeev, K. A. Venkatesan, K. Ananthasivan 
23 Nov 2022-Radiochimica Acta
TL;DR: In this paper , the effectiveness of the solvent wash reagents such as sodium carbonate and hydrazine carbonate were compared using the lean solvent obtained during reprocessing of high plutonium content fast reactor spent nuclear fuel.
Abstract: Abstract The effectiveness of the solvent wash reagents such as sodium carbonate and hydrazine carbonate were compared using the lean solvent obtained during reprocessing of high plutonium content fast reactor spent nuclear fuel. The concentrations of the reagents were optimized based on the removal of retained metal ions from the lean solvent and 1.5 M hydrazine carbonate was found to be optimal and superior to sodium carbonate to be adopted as the reagent for primary cleanup of the spent solvent. Both the reagents were found to possess excellent room temperature stability and their performance was not affected upon gamma irradiation which was assessed in terms of capacity to remove di-butyl phosphate from degraded TBP/n-DD system.

1 citations

Ion Exchange in Nuclear Fuel Reprocessing : Zirconium Phosphate Materials for the Separation of Trivalent Actinides and Lanthanides

[...]

Elmo Wiikinkoski
29 Nov 2019
TL;DR: In this paper, the authors present a list of original publications and their corresponding abstracts, including acknowledgements, references, and abbreviations of the original publications, and references.
Abstract: .................................................................................................................1 Acknowledgements ................................................................................................2 List of original publications .....................................................................................3 Abbreviations .........................................................................................................4

Cites methods from "Macroreticular ion exchange resin c..."

  • ...Anion exchange resins were used for the clean-up of PUREX waste as they readily removed the fission products and TBP degradation products in the waste.[19,20] Historically most importantly, both cation and anion exchange have been used for concentration and purification for plutonium from the...

    [...]

Journal Article•DOI•
Solvent wash studies for the removal of di-butyl phosphate from spent solvent under simulated PUREX condition

[...]

Satyabrata Mishra, P. Vishnu Anand, Chayan Patra, Pranay Kumar Sinha, C. K. Mukhopadhyay, Jammu Ravi, Dandapani Sivakumar, Narasimhan Desigan, Puspalata Rajesh, R. Rajeev, K. Dhamodharan, P. Velavendan, K. A. Venkatesan, K. Ananthasivan 
10 Jan 2023-Journal of Radioanalytical and Nuclear Chemistry
Related Papers (5)
Clean Up of the Purex Process TBP Solvent by Macroreticular Ion Exchange Resin

[...]

01 Feb 1975-Radiochimica Acta
Μ. Κ. Rahman
Macroreticular anion exchange resin cleanup of tbp solvents.

[...]

01 Jan 1972-Transactions of the American Nuclear Society
W.W. Schulz
Thermochemical measurements of purex process solutions

[...]

01 May 1956
G.A. Burney
A stable diluent for purex process extractants

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01 Jun 1958
H. Pollock
The extraction of cesium and strontium from acidic high activity nuclear waste using a purex process compatible organic extractant

[...]

01 Oct 1985-Solvent Extraction and Ion Exchange
R. G. Shuler, C. B. Bowers, J. E. Smith, V. Van Brunt, M.W. Davis