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Title: MST-7 Engineered Materials

Abstract

Materials design, fabrication, assembly, and characterization for national security needs

Authors:
 [1]
  1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1352357
Report Number(s):
LA-UR-17-22799
DOE Contract Number:
AC52-06NA25396
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE

Citation Formats

Kippen, Karen Elizabeth. MST-7 Engineered Materials. United States: N. p., 2017. Web. doi:10.2172/1352357.
Kippen, Karen Elizabeth. MST-7 Engineered Materials. United States. doi:10.2172/1352357.
Kippen, Karen Elizabeth. Thu . "MST-7 Engineered Materials". United States. doi:10.2172/1352357. https://www.osti.gov/servlets/purl/1352357.
@article{osti_1352357,
title = {MST-7 Engineered Materials},
author = {Kippen, Karen Elizabeth},
abstractNote = {Materials design, fabrication, assembly, and characterization for national security needs},
doi = {10.2172/1352357},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Apr 06 00:00:00 EDT 2017},
month = {Thu Apr 06 00:00:00 EDT 2017}
}

Technical Report:

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  • Engineered forms of MST and mMST were prepared at ORNL using an internal gelation process. Samples of these two materials were characterized at SRNL to examine particle size and morphology, peroxide content, tapped densities, and Na, Ti, and C content. Batch contact tests were also performed to examine the performance of the materials. The {sup E}mMST material was found to contain less than 10% of the peroxide found in a freshly prepared batch of mMST. This was also evidenced in batch contact testing with both simulated and actual waste, where little difference in performance was seen between the two engineeredmore » materials, {sup E}MST and {sup E}mMST. Based on these results, attempts were made to increase the peroxide content of the materials by post-treatment with hydrogen peroxide. The peroxide treatment resulted in a slight ({approx}10%) increase in peroxide content; however, the peroxide:Ti molar ratio was still much lower ({approx}0.1 X) than what is seen in a freshly prepared batch of mMST. Testing with simulated waste showed the performance of the peroxide treated materials was improved. Batch contact tests were also performed with an earlier (2003) prepared lot of {sup E}MST to examine the effect of ionic strength on the performance of the material. In general the results showed a decrease in removal performance with increasing ionic strength, which is consistent with previous testing with MST. A Sr loading isotherm was also determined, and the {sup E}MST material was found to reach a Sr loading as high as 13.2 wt % after 100 days of contact at a phase ratio of 20000 mL/g. At the typical MST phase ratio of 2500 mL/g (0.4 g/L), a Sr loading of 2.64 wt % was reached after 506 hours of contact. Samples of {sup E}MST and the post-peroxide treated {sup E}mMST were also tested in a column configuration using simulated waste solution. The breakthrough curves along with analysis of the sorbent beds at the conclusion of the experiments showed that the peroxide treated {sup E}mMST has a higher Sr and Np capacity, but that both materials have similar Pu capacities. The {sup E}MST removed a larger percentage of U than the peroxide treated {sup E}mMST, which is consistent with previous testing which showed that mMST has little affinity for U under these conditions.« less
  • A process for the recovery of plutonium and americium from molten salt extraction (MSE) salt residues has been demonstrated. It is based upon a new chloride anion-exchange process at low acidity that eliminates corrosive HCl fumes. The Los Alamos americium oxide production line has been improved to give more product with a concurrent lowering of personnel radiation exposure. A cost study has been made for the disposal of americium-contaminated calcium metal buttons that were obtained by pyrochemical recovery of plutonium from MSE salts. The waste form used in the study conforms to WIPP-Facility standards and current state-of-the-art radioactive waste disposal.more » The cost estimate is approx. $300/g /sup 241/Am. Plutonium decontamination factors of approx. 300 have been obtained from lead-platinum alloy dissolution experiments carried out in alumina crucibles using lead oxide slag to getter the plutonium.« less
  • Savannah River National Laboratory (SRNL) performed experiments on qualification material for use in the Interim Salt Disposition Program (ISDP) Batch 7 processing. The Marcrobatch 7 material was received with visible fine particulate solids, atypical for these samples. The as received material was allowed to settle for a period greater than 24 hours. The supernatant was then decanted and utilized as our clarified feed material. As part of this qualification work, SRNL performed an Actinide Removal Process (ARP) test using the clarified feed material. From this test, the residual monosodium titanate (MST) was analyzed for radionuclide uptake after filtration from H-Tankmore » Farm (HTF) feed salt solution. The results of these analyses are reported and are within historical precedent.« less