FISSION PRODUCT TRANSPORT THROUGH GRAPHITE MATRICES

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Description

The transport of fission products from points of origin in unclad graphite matrix-type fuels to the reactor circulating system involves, as one of the steps. diffusion through the graphite matrix to the fuel element surface. As pointed out by Rosenthal. the fraction of a given fission product chain actaally reaching the fuel element surface will be small if the time for transport through the graphite is long compared to the half-lives of the volatile members. An important problem, therefore, is the determination of the effective transport rates of the various mobile elements and their daughter products of interest through various ... continued below

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Pages: 24

Creation Information

Korsmeyer, R.B. March 21, 1961.

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Description

The transport of fission products from points of origin in unclad graphite matrix-type fuels to the reactor circulating system involves, as one of the steps. diffusion through the graphite matrix to the fuel element surface. As pointed out by Rosenthal. the fraction of a given fission product chain actaally reaching the fuel element surface will be small if the time for transport through the graphite is long compared to the half-lives of the volatile members. An important problem, therefore, is the determination of the effective transport rates of the various mobile elements and their daughter products of interest through various graphites suitable for use as fuel element compacts. as functions of temperature over the range of greatest immediate interest to reactor designers. The upper end of the range need not exceed about 1000 deg C. The transport of helium and arbon through various graphites has been the subject of considerable study by Watson. Evans, and other,. and a prelimilnary investigation of the high temperature transport of some ordinarily non-volatile elements has been carried out by Saunders. This work is briefly reviewed in relation to the final problem and the areas in which further information is needed most by reactor designers is indicated. (auth)

Physical Description

Pages: 24

Source

  • Other Information: Orig. Receipt Date: 31-DEC-61

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  • Report No.: CF-61-3-75
  • Grant Number: None
  • DOI: 10.2172/4055679 | External Link
  • Office of Scientific & Technical Information Report Number: 4055679
  • Archival Resource Key: ark:/67531/metadc872458

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Creation Date

  • March 21, 1961

Added to The UNT Digital Library

  • Sept. 16, 2016, 12:32 a.m.

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  • Nov. 28, 2016, 4:37 p.m.

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Korsmeyer, R.B. FISSION PRODUCT TRANSPORT THROUGH GRAPHITE MATRICES, report, March 21, 1961; Tennessee. (digital.library.unt.edu/ark:/67531/metadc872458/: accessed September 24, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.