skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Direct LiT Electrolysis in a Metallic Fusion Blanket

Abstract

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

Authors:
 [1]
  1. Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
Publication Date:
Research Org.:
Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
Sponsoring Org.:
USDOE Office of Environmental Management (EM)
OSTI Identifier:
1327787
Report Number(s):
SRNL-STI-2016-00572
TRN: US1700320
DOE Contract Number:
AC09-08SR22470
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CHEMISTRY; 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; TRITIUM; LITHIUM; IONIC CONDUCTIVITY; ELECTROLYSIS; TRITIUM RECOVERY; LITHIUM IONS; LITHIUM TRITIDES; CONCENTRATION RATIO; TEMPERATURE RANGE 0400-1000 K; DEUTERIUM; HYDROGEN; BREEDING BLANKETS; CERAMICS; HYBRID REACTORS; EXTRACTION

Citation Formats

Olson, Luke. Direct LiT Electrolysis in a Metallic Fusion Blanket. United States: N. p., 2016. Web. doi:10.2172/1327787.
Olson, Luke. Direct LiT Electrolysis in a Metallic Fusion Blanket. United States. doi:10.2172/1327787.
Olson, Luke. 2016. "Direct LiT Electrolysis in a Metallic Fusion Blanket". United States. doi:10.2172/1327787. https://www.osti.gov/servlets/purl/1327787.
@article{osti_1327787,
title = {Direct LiT Electrolysis in a Metallic Fusion Blanket},
author = {Olson, Luke},
abstractNote = {A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.},
doi = {10.2172/1327787},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

Save / Share:
  • A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed.  The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This workmore » identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. « less
  • The release and transport of activated materials-of-construction in a fusion reactor during an accident scenario involving overheating and ingress of oxidants is an important area of safety research. This investigation quantified material release characteristics that result from surface oxide spallation and vaporization from the steel alloys PCA and HT-9 in impure helium and air environments. Flowing air and helium, each containing specific quantities of O/sub 2/ and H/sub 2/O, were used to oxidize test sample surfaces at temperatures of 800/sup 0/ and 1000/sup 0/C for exposure times of <200 h. The changes and features observed are described and include: weight,more » oxide scale morphology, adherence and composition; alloy composition (including decarburization); and vaporization as fractional loss of alloying elements. Oxide scales formed were dominant in Mn and Cr but minor in Fe. The dominant volatilized elements detected were Mo, W, Cr, As, Mn, Sb, and Co. The implications of these data for safety analyses of activated material release are that following an accidental temperature excursion to 800/sup 0/C in an oxidizing environment, material transport by scale spallation and/or volatilization should be minor. Furthermore, the potential material release fractions at 1000/sup 0/C appear to be significantly lower than the release fractions used in early fusion safety analyses.« less
  • In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components.