Document Details


Title:
Thorex Pilot Plant Run HD-19 Summary
Subject Terms:
ORNL Thorex Pilot Plant; extraction, Pa-233, U-233; fission products; flowsheet, HD-19 summary; slug, decontamination; thorium, metal, iodine, I-131; uranium, ruthenium, solvent
Addressee:
Culler, F.L.
Document Location:
DOE INFORMATION CENTER 1 Science.gov Way, Oak Ridge, TN 37831; Eva Butler; Phone: 865-241-4780; Toll-Free: 800-382-6938, Option 6; FAX: 865-574-3521; Email: doeic@oro.doe.gov
Document Categories:
Health, Safety and Environment\Public Health and Safety; Health, Safety and Environment\Public Health and Safety
Document Type:
REPORT
Publication Date:
1957 Apr 30
Declassification Date:
1977 Nov 18
Declassification Status:
Declassified
Document Pages:
80
Accession Number:
ORF65765
Document Number(s):
CF-57-4-1
Originating Research Org.:
Union Carbide Nuclear Division
OpenNet Entry Date:
1998 Jun 16
Description/Abstract:
This document is concerning the Thorex Pilot Plant Run HD-19 summary. This plant was designed and constructed to develop and demonstrate a single-thorium-cycle solvent extraction process for the separation of thorium, U-233, and fission products (including Pa-233) from reactor-irradiated thorium metal and to recover the thorium and uranium as aqueous solutions suitable for direct handling in subsequent processing. It became desirable to determine the chemical suitability of the ORNL Thorex flowsheet for processing such feeds and to define the problems associated with such processing. Based on the results of run HD-19 and of the two previous production runs, the one-cycle Thorex flowsheet (ORNL Thorex flowsheet No. 1.2-7) provided insufficient decontamination from Pa-233 and fission products when thorium irradiated to greater than ~1500 g/t and decayed ~30 days or less is processed. More than 90 percent of the contamination of the thorium and uranium products was due to ruthenium. The tributyl phosphate-Amsco diluent mixture used as an extractant (solvent) was degraded to a greater degree than previously experienced in any run upon exposure to the process solutions and in the radiation fields associated with processing 3340 g/t, 109-day-decayed or 2260 g/t, 29-day-decayed thorium. Iodine contamination of the solvent increased by a factor of ~700 when 2260 g/t, 29-day-decayed feed was processed as compared to the 3340 g/t, 109-day-decayed thorium feed. Sodium carbonate scrubbing was not effective in removing iodine from iodine-contaminated solvent; iodine decontamination factors of 1-1.5 were obtained. The total uranium recovery in run HD-19 was 1794 g. Periodic purging of the solids-containing emulsion belts from the solvent-extraction column liquid-liquid interfaces reduced entrainment of highly contaminated particles in the solvent streams leaving the column. Recycling aqueous solutions to the extraction columns concurrently with adjusted feed was mechanically successful at flowrates about 30 to 50 percent of that of the adjusted feed. Thorium losses were about 30 percent higher during such periods. Processing 3340 g/t, 109-day-decayed or 2260 g/t, 29-day-decayed thorium was mechanically successful. The quality of the solvent deteriorated rapidly as the run progressed, becoming more pronounced as the 29-day decayed thorium was processed. Tables and figures are included.


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