Diffusion of Fission Fragments From Uranium Impregnated Graphite

Smith, Charlyne; Young, Clyde T.

doi:10.2172/4325862

Cited by 2 publications

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“…approximate diffusion coefficients f o r DiffusionCoefficients for Xenon and Krypton in U BOOTH AND RYMER -----------CRDC -7 2 0SUSKO -----------------------GENEVA PAPER 2 4 0 s i o n products out of AGOT graphit.e from the data i n Refs (5). and(10)(11)(12)(13)). Thesec o e f f i c i e n t s m a y be used to.compute agproximate release r a t e s f o r f i s s i o n products out of NOT graphite* The r a t i o of t h e releqse of a c t i v i t y from the f u e l by r e c o i l t o t h a t by d i f f u s i o n f r o n the f u e l depends on both the diffusion process which i s temperat u r e dependent and the p a r t i c l e s i z e of the fuel, which i n turn depends on the method of fabrication.…”

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confidence: 99%

A Study of Problems Associated With Release of Fission Products From Ceramic Fuels in Gas-Cooled Reactors

Lane¹,

Bennett²,

Culver³

et al. 1959

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This repat was prepared as an o~counS of Govomarscrc sponsored work. Weirher th. United Stutee, not the Comdission, nor. on); poreom ocflng on b6half of the Commlssionr A. h k r s any wurrmty or reprbsentotion, expressed or impllud, wRh rropect t o tbr accufflcy. comp4rt.mrs, or usefulness of the iniarmation conmlned i n fhis reporf, or h a t fhe u r e of any inferination, oppmrotuc, method, or procass &isclosed in thh fupori may net infringe privately owned rights: or B. Aasumoc any Ilobillties with t e s p c t to the use of, or for domagn resulting from the use of any informbatio~, oppo?stur. method, or promess disclosed in this repMa As wed I n the above, "person acting on behalf of the Commis~ion'~ includes m y ehtployee or sonnacto, of the Commircfon, or employee of sweh centmotor, to the extent that sucrh e m~i o p e or contractor of the Commission, or employee of such contractor preporor, dimeminotea. or provides occess to, any information purswnt t o his employment oz contract with the CommissEon, or his employment with such controetar.

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mentioning

confidence: 99%

A Study of Problems Associated With Release of Fission Products From Ceramic Fuels in Gas-Cooled Reactors

Lane¹,

Bennett²,

Culver³

et al. 1959

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Graphite as a Fuel Matrix

Riley

1960

Materials in Nuclear Applications

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Recent work on fabrication development, laboratory evaluation, and reactor evaluation of fueled graphite is reviewed. Methods of preparing fueled graphites are discussed with respect to possible differences in reactor perlonnance and processing costs. Geometries and fuel loadings that have been achieved are mentioned. Data are given on variation in physical properties with fuel content. The experimental evidence presented indicates that neutron plus fission fragment irradiation at temperatures of about 1300 F tends to cause the matrix structure to change from graphite to disordered carbon. With respect to dimensional distortion, the individual effects of fission fragment and neutron irradiation at about 1300 F have been separated. Each appears to cause contraction. A relationship between dimensional distortion and the number ol fission fragments attenuated in the matrix is suggested. Fission product release from a number of types of fueled graphites are reported. Results of both laboratory heat-treatment studies following low-temperature, low-exposure irradiation and high-temperature in-pile studies indicated that the release of fission gases was the order of at least 1 per cent of those calculated to be present. Methods for improving fission product retention are discussed. The experimental evidence presented indicates that surlace coatings, impervious graphite, and impervious cladding for fuel particles have some potential. Available data on chemical reactions between fuel compounds and the graphite matrix are discussed briefly. Recommendations for future developmental work are included. Emphasis should be placed on improved fission product retention and studies of radiation effects at high exposures.

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Diffusion of Fission Fragments From Uranium Impregnated Graphite

Cited by 2 publications

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A Study of Problems Associated With Release of Fission Products From Ceramic Fuels in Gas-Cooled Reactors

A Study of Problems Associated With Release of Fission Products From Ceramic Fuels in Gas-Cooled Reactors

Graphite as a Fuel Matrix

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