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Title: The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

Abstract

The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest.

Authors:
;  [1]
  1. Argonne National Lab., IL (United States). Technology Development Div.
Publication Date:
OSTI Identifier:
413388
Resource Type:
Journal Article
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 116; Journal Issue: 2; Other Information: PBD: Nov 1996
Country of Publication:
United States
Language:
English
Subject:
05 NUCLEAR FUELS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; SPENT FUELS; PYROCHEMICAL REPROCESSING; G CODES; MATHEMATICAL MODELS; WATER COOLED REACTORS; MIXED OXIDE FUELS

Citation Formats

Ahluwalia, R K, and Geyer, H K. The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels. United States: N. p., 1996. Web.
Ahluwalia, R K, & Geyer, H K. The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels. United States.
Ahluwalia, R K, and Geyer, H K. 1996. "The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels". United States.
@article{osti_413388,
title = {The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels},
author = {Ahluwalia, R K and Geyer, H K},
abstractNote = {The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest.},
doi = {},
url = {https://www.osti.gov/biblio/413388}, journal = {Nuclear Technology},
number = 2,
volume = 116,
place = {United States},
year = {Fri Nov 01 00:00:00 EST 1996},
month = {Fri Nov 01 00:00:00 EST 1996}
}