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Title: Laboratory studies in support of the West Valley sludge mobilization wash

Abstract

The vitrification of the West Valley Demonstration Project (WVDP) high-level waste (HLW) requires the removal of substantial amounts of sulfate. The solubility of sulfate in the glass produced in the existing flow sheet using a slurry-fed ceramic melter is 0.3 wt%. Any sulfate above this level would form a molten salt layer on the surface of the melt. This layer would interfere with the melting process and result in an unacceptable borosilicate glass. The amount of sludge that could be put in the glass would be much less than planned, and the number of glass canisters necessary for solidifying the WVDP HLW would be substantially increased. Laboratory studies were recently performed to determine an acceptable flow sheet and operational parameters for the sulfate removal. Three technologies that meet the WVDP requirements were identified and tested in the laboratory to reduce the concentration of plutonium and uranium in the wash water. At this time, experimental results strongly indicate that the technologies can be implemented successfully at the WVDP. Current information also suggests that there should be no significant delay in solidifying the HLW as a result of using these methods.

Authors:
; ; ; ;  [1]
  1. West Valley Nuclear Services Co., NY (United States)
Publication Date:
OSTI Identifier:
7194760
Report Number(s):
CONF-920606-
Journal ID: ISSN 0003-018X; CODEN: TANSA
Resource Type:
Conference
Journal Name:
Transactions of the American Nuclear Society; (United States)
Additional Journal Information:
Journal Volume: 65; Conference: American Nuclear Society annual meeting, Boston, MA (United States), 7-12 Jun 1992; Journal ID: ISSN 0003-018X
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; HIGH-LEVEL RADIOACTIVE WASTES; VITRIFICATION; WEST VALLEY PROCESSING PLANT; RADIOACTIVE WASTE MANAGEMENT; BOROSILICATE GLASS; CERAMIC MELTERS; CESIUM; CHEMICAL REACTIONS; DEMONSTRATION PROGRAMS; ECONOMICS; ION EXCHANGE; LOW-LEVEL RADIOACTIVE WASTES; MELTING; PH VALUE; PLUTONIUM RECYCLE; REMOVAL; REPROCESSING; SLUDGES; SPENT FUELS; STRONTIUM; SULFATES; TEMPERATURE DEPENDENCE; TITANIUM; UNDERGROUND STORAGE; URANIUM RECYCLE; WASTE RETRIEVAL; ALKALI METALS; ALKALINE EARTH METALS; ELECTRIC FURNACES; ELEMENTS; ENERGY SOURCES; FUEL CYCLE; FUEL REPROCESSING PLANTS; FUELS; FURNACES; GLASS; MANAGEMENT; MATERIALS; METALS; NUCLEAR FACILITIES; NUCLEAR FUELS; OXYGEN COMPOUNDS; PHASE TRANSFORMATIONS; RADIOACTIVE MATERIALS; RADIOACTIVE WASTES; REACTOR MATERIALS; SEPARATION PROCESSES; STORAGE; SULFUR COMPOUNDS; TRANSITION ELEMENTS; WASTE MANAGEMENT; WASTES; 052001* - Nuclear Fuels- Waste Processing

Citation Formats

Fauth, D J, Michnik, L E, Palmer, R A, Hara, F T, and Kazmierczak, T F. Laboratory studies in support of the West Valley sludge mobilization wash. United States: N. p., 1992. Web.
Fauth, D J, Michnik, L E, Palmer, R A, Hara, F T, & Kazmierczak, T F. Laboratory studies in support of the West Valley sludge mobilization wash. United States.
Fauth, D J, Michnik, L E, Palmer, R A, Hara, F T, and Kazmierczak, T F. 1992. "Laboratory studies in support of the West Valley sludge mobilization wash". United States.
@article{osti_7194760,
title = {Laboratory studies in support of the West Valley sludge mobilization wash},
author = {Fauth, D J and Michnik, L E and Palmer, R A and Hara, F T and Kazmierczak, T F},
abstractNote = {The vitrification of the West Valley Demonstration Project (WVDP) high-level waste (HLW) requires the removal of substantial amounts of sulfate. The solubility of sulfate in the glass produced in the existing flow sheet using a slurry-fed ceramic melter is 0.3 wt%. Any sulfate above this level would form a molten salt layer on the surface of the melt. This layer would interfere with the melting process and result in an unacceptable borosilicate glass. The amount of sludge that could be put in the glass would be much less than planned, and the number of glass canisters necessary for solidifying the WVDP HLW would be substantially increased. Laboratory studies were recently performed to determine an acceptable flow sheet and operational parameters for the sulfate removal. Three technologies that meet the WVDP requirements were identified and tested in the laboratory to reduce the concentration of plutonium and uranium in the wash water. At this time, experimental results strongly indicate that the technologies can be implemented successfully at the WVDP. Current information also suggests that there should be no significant delay in solidifying the HLW as a result of using these methods.},
doi = {},
url = {https://www.osti.gov/biblio/7194760}, journal = {Transactions of the American Nuclear Society; (United States)},
issn = {0003-018X},
number = ,
volume = 65,
place = {United States},
year = {Wed Jan 01 00:00:00 EST 1992},
month = {Wed Jan 01 00:00:00 EST 1992}
}

Conference:
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