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Title: HOT CELL DEMONSTRATION OF ZIRFLEX AND SULFEX PROCESSES. Report No. 3

Abstract

ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiatedmore » to levels greater than 12,000 Mwd/T causing a decrease in Zr Z'' test by a factor of 15. (nuth)« less

Authors:
;
Publication Date:
Research Org.:
Oak Ridge National Lab., Tenn.
OSTI Identifier:
4819023
Report Number(s):
ORNL-TM-187
NSA Number:
NSA-16-027123
DOE Contract Number:
W-7405-ENG-26
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-62
Country of Publication:
United States
Language:
English
Subject:
CHEMISTRY; AMSCO; BUTYL PHOSPHATES; CENTRIFUGATION; COATING; DECOMPOSITION; DECONTAMINATION; DENSITY; EFFICIENCY; EXTRACTION COLUMNS; FUEL CANS; GAMMA RADIATION; HOT CELLS; IRRADIATION; KEROSENE; LOSSES; NITRIC ACID; ORGANIC NITROGEN COMPOUNDS; OXALIC ACID; PELLETS; PLUTONIUM COMPOUNDS; POWDERS; PUREX PROCESS; RADIATION CHEMISTRY; RADIATION DOSES; RADIATION PROTECTION; REMOTE HANDLING; REPROCESSING; SEPARATION PROCESSES; SOLUBILITY; SOLUTIONS; SOLVENT EXTRACTION; SOLVENTS; SULFEX PROCESS; TBP; URANIUM COMPOUNDS; URANIUM DIOXIDE; WEAR; ZIRCALOY; ZIRCONIUM COMPOUNDS; ZIRFLEX PROCESS

Citation Formats

Goode, J.H., and Baillie, M.G. HOT CELL DEMONSTRATION OF ZIRFLEX AND SULFEX PROCESSES. Report No. 3. United States: N. p., 1962. Web. doi:10.2172/4819023.
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Goode, J.H., & Baillie, M.G. HOT CELL DEMONSTRATION OF ZIRFLEX AND SULFEX PROCESSES. Report No. 3. United States. doi:10.2172/4819023.
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Goode, J.H., and Baillie, M.G. Mon . "HOT CELL DEMONSTRATION OF ZIRFLEX AND SULFEX PROCESSES. Report No. 3". United States. doi:10.2172/4819023. https://www.osti.gov/servlets/purl/4819023.
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@article{osti_4819023,
title = {HOT CELL DEMONSTRATION OF ZIRFLEX AND SULFEX PROCESSES. Report No. 3},
author = {Goode, J.H. and Baillie, M.G.},
abstractNote = {ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiated to levels greater than 12,000 Mwd/T causing a decrease in Zr Z'' test by a factor of 15. (nuth)},
doi = {10.2172/4819023},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon May 14 00:00:00 EDT 1962},
month = {Mon May 14 00:00:00 EDT 1962}
}
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DOI:  10.2172/4819023
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  • ;
    Twelve Zircaloy-2 clad UO/sub 2/ specimens, irradiated in the NRK to about 15,500 Mwd/t and decayed about 9 months, were declad and core dissolutions completed. The pellets were fractured brt were essentially intact (i.e. had not fallen apart) when declad. The U losses ramged from 0.03 to 0.09%, due to solubility of U(IV) in the Zinflex reagent, and the U losses from 0.01 to 0.04%. The core pellets were 99.5% dlssolved in 4 M HNO/sub 3/-0.1M Al(NO/sub 3/)/sub 3/ in 5 hr, which was slightly faster than the rate of unirradiated pellets. (auth)

  • ;
    Four runs were conducted in the Zirflex-Sulfex headend hot cell equipment. Prototype PWR blanket rods, Zircaloy2 clad UO/sub 2/, irradiated from 159 to 356 Mwd/t and decayed 2 years, were declad in boiling 6 M NH/sub 4/F-l M NH/sub 4/NO/sub 3/, terminating with a F/Zr mol ratio of 7 in the spent decladding solution. Average decladding time was 1.5 hr, leaving end cap residues of about 5 g per pin. At the end of the decladding, maximum loss of uranium and plutonium to the decladding solution was 0.04 and 0.37%, respectively. The core pellets were largely shattered with less thanmore » 0.5 wt% smaller than 10 mesh. Core dissolution was complete in 5 M HNO/sub 3/ in about 40 minutes, yielding a solvent extraction feed containing 4 M HNO/sub 3/ and 100 g U/l. Solid residue from the decladding and core dissolution was less than 0.001% of the initial weight and consisted of traces of Ca, Fe, Cr, and Sn; uranium and plutonium were not detected. (auth)« less

  • ; ; - Transactions of the American Nuclear Society (U.S.)

  • ; ;
    The Zirflex and Sulfex processes for chemical decladding Zircaloy or stainless-steel-clad UO/sub 2/ power reactor fuels were successfully demonstrated at irradiation levels as high as 28,200 Mwd/t. The Zircaloy jackets were dissolved in boiling 6 M NH/sub 4/F-- 1 M NH/sub 4/NO/sub 3/, and the stainless steel jackets were dissolved in refluxing 4 M H/sub 2/SO/sub 4/. Both processes gave average soluble losses of uranium and plutonium to the decladding reagents of about 0.05%. Centrifugation or filtration of the highly radioactive decladding waste solutions was required to recover UO/sub 2/ fines produced by fracture of the UO/sub 2/. The finesmore » were recycled and dissolved with the UO/sub 2/ cores in boiling 4 M HNO/sub 3/ solution. About 5 to 6 hr were required for complete dissolution of the UO/sub 2/ core to produce terminal concentrations of 100 g of uranium per liter and 3 M HNO/sub 3/. The core solution was a suitable solvent extraction feed after clarification and adjustment of plutonium valence with sodium nitrite. One cycle of the modified Purex process, in Mini mixersettlers, using 100-g-uranium-per-liter feed solutions, gave losses of uranium and plutonium to the raffinate of less than 0.1%, and gross gamma decontamination factors of about 1.5 x 10/sup 4/ for the uranium product and about 5 x 10/sup 3/ for the plutonium product. High purity n-dodecane diluent for the 30% TBP solvent retarded the formation of nonremovable uranium-retaining degradation products in the solvent by factors of up to 20, compared to Amsco- diluted TBP as measured by the uranium content of the washed solvent after six simulated cycles through the modified Purex process. The feed solutions were prepared from fuel specimens irradiated to 13,000 Mwd/t and cooled about 1 year. (auth)« less

  • In order to recover uranium from zirconium-base reactor fuels by solvent extraction, the metailic fuel and cladding must first be dissolved and a suitable feed solution prepared. Such preparations of solvent extraction feeds were successfully accomplished batchwise using both the Modified Zirflex and Neuflex processes employing an NH/sub 4/F -- oxidant mixture to dissolve the fuel elements, and the feed. (The d Zirflex feed, and H/sub 2/O for the Neuflex feed.) In the Modified Zirflex process, a dissolvent about 6 M in NH/sub 4/F with an excess of H/sub 2/O/sub 2/ to oxidize uranium to the more-soluble U(VI) valence state.more » The off-gas, after NH/sub 3/ removal, is an H/sub 2/-O/sub 2/ mixture of small volume, which is diluted with air to a safe concentration. Then nitric acid-aluminum nitrate is added to the dissolution product, yielding a solvent extraction feed from which uranium is recovered by using TBP-Amsco as the extractant. In the Neuflex process, the dissolvent is NH/sub 4/F--H/sub 2/O/sub 2/, with less than a stoichiometric amount of NH/sub 4/NO/sub 3/. Without NH/sub 4/NO/sub 3/, the scrubbed off-gas is principally hydrogen, on the hydrogen-rich side of the flammable range of H/sub 2/-O/sub 2/ mixtures, Only water is added to this dissolution product, yielding a neutral fluoride feed from which uranium is extractable by use of Dapex reagents. ln both processes the F: Zr charge ratio, initial surface condition, and maximum section thickness of the fuel element were the principa1 determinants of total dissolution time. The zirconium loading as determined by the free fluoride - zirconium solubility relationship limited the capacity of fuels containing less than 2% U, while the free-fluoride-to-uranium ratio of about 100 required for solution stability was the limiting factor with alloys containing higher percentages of uranium, Hydrogen peroxide concentration was not an important factor in solution stability; the role of ammonla or NH/sub 4/OH was not studied. The feasibility of both processes was demonstrated by a series of batch dissolutions of kilogram quartities of various fuels containing 1 to 8% uranium. Continuous dissolution was demonstrated as was application to TRIGA fuel alloy (8% U-- ZrH). Stainless steel type 347 and a low-carbon nickel alloy were suitable materials of construction for the dissolution and the solvent extraction equipment. Since there were some discrepancies betweeq small-scale and engineering-scale work, especially in the prevention of precipitate formation near the end of the dissolution cycle, it is advised that some further investigation be made prior to attempted scaleup to plant operation. (auth)« less