Tritium recovery from some ceramic breeding materials

H. Kwast; A. Kout; Ruud P. Muis; Marcel P. Stijkel; A. J. Flipot; J. D. Elen

doi:10.1016/S0920-3796(89)80132-2

Tritium recovery from some ceramic breeding materials

H. Kwast, A. Kout, Ruud P. Muis, Marcel P. Stijkel, A. J. Flipot, J. D. Elen

Research output › peer-review

Abstract

Tritium retention measurements have been carried out on samples of LiAlO₂, Li₂SiO₃ and Li₂O of various density, irradiation time and temperature. The reported results were obtained by dissolution of the samples in concentrated KOH and by heating the samples up to 900 °C under a helium flow. Both techniques are described. The results are compared with the in-situ tritium release experiments carried out in the High Flux Reactor at Petten, The Netherlands. The in-situ release and post-irradiation retention measurements are in good agreement. Tritium inventory depends predominantly on temperature and density. For the materials with a density of approximately 80% TD and irradiated at about 600 °C tritium inventories of ≪1 wppm can be achieved. Higher densities and lower temperatures result in inventories > 1 wppm.

Original language	English
Pages (from-to)	359-364
Number of pages	6
Journal	fusion engineering and design
Volume	8
Issue number	C
DOIs	https://doi.org/10.1016/S0920-3796(89)80132-2
State	Published - 1989

ASJC Scopus subject areas

Civil and Structural Engineering
Nuclear Energy and Engineering
General Materials Science
Mechanical Engineering

Access to Document

10.1016/S0920-3796(89)80132-2

Cite this

@article{29403ce68bfb45f68f4ce377f787be60,

title = "Tritium recovery from some ceramic breeding materials",

abstract = "Tritium retention measurements have been carried out on samples of LiAlO2, Li2SiO3 and Li2O of various density, irradiation time and temperature. The reported results were obtained by dissolution of the samples in concentrated KOH and by heating the samples up to 900 °C under a helium flow. Both techniques are described. The results are compared with the in-situ tritium release experiments carried out in the High Flux Reactor at Petten, The Netherlands. The in-situ release and post-irradiation retention measurements are in good agreement. Tritium inventory depends predominantly on temperature and density. For the materials with a density of approximately 80% TD and irradiated at about 600 °C tritium inventories of ≪1 wppm can be achieved. Higher densities and lower temperatures result in inventories > 1 wppm.",

author = "H. Kwast and A. Kout and Muis, {Ruud P.} and Stijkel, {Marcel P.} and Flipot, {A. J.} and Elen, {J. D.}",

year = "1989",

doi = "10.1016/S0920-3796(89)80132-2",

language = "English",

volume = "8",

pages = "359--364",

journal = "fusion engineering and design",

issn = "0920-3796",

publisher = "Elsevier",

number = "C",

}

TY - JOUR

T1 - Tritium recovery from some ceramic breeding materials

AU - Kwast, H.

AU - Kout, A.

AU - Muis, Ruud P.

AU - Stijkel, Marcel P.

AU - Flipot, A. J.

AU - Elen, J. D.

PY - 1989

Y1 - 1989

N2 - Tritium retention measurements have been carried out on samples of LiAlO2, Li2SiO3 and Li2O of various density, irradiation time and temperature. The reported results were obtained by dissolution of the samples in concentrated KOH and by heating the samples up to 900 °C under a helium flow. Both techniques are described. The results are compared with the in-situ tritium release experiments carried out in the High Flux Reactor at Petten, The Netherlands. The in-situ release and post-irradiation retention measurements are in good agreement. Tritium inventory depends predominantly on temperature and density. For the materials with a density of approximately 80% TD and irradiated at about 600 °C tritium inventories of ≪1 wppm can be achieved. Higher densities and lower temperatures result in inventories > 1 wppm.

AB - Tritium retention measurements have been carried out on samples of LiAlO2, Li2SiO3 and Li2O of various density, irradiation time and temperature. The reported results were obtained by dissolution of the samples in concentrated KOH and by heating the samples up to 900 °C under a helium flow. Both techniques are described. The results are compared with the in-situ tritium release experiments carried out in the High Flux Reactor at Petten, The Netherlands. The in-situ release and post-irradiation retention measurements are in good agreement. Tritium inventory depends predominantly on temperature and density. For the materials with a density of approximately 80% TD and irradiated at about 600 °C tritium inventories of ≪1 wppm can be achieved. Higher densities and lower temperatures result in inventories > 1 wppm.

UR - http://www.scopus.com/inward/record.url?scp=0040235999&partnerID=8YFLogxK

U2 - 10.1016/S0920-3796(89)80132-2

DO - 10.1016/S0920-3796(89)80132-2

M3 - Article

AN - SCOPUS:0040235999

SN - 0920-3796

VL - 8

SP - 359

EP - 364

JO - fusion engineering and design

JF - fusion engineering and design

IS - C

ER -