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Title: Fuel element development program for the Pebble Bed Reactor. Final Report

Abstract

The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found that molecularly deposited ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces and where a larger thickness-todiameter ratio could be used than if the coating were on the surface of the graphite sphere. Fuel elements were irradiated to burnups ranging up to about 6 at.% U/sup 235/. In all specimens containing a uniform dispersion of fuel, the graphite spheres were found to retain their structural properties after irradiation. Data are given on fuel particle coatings of A1/sub 2/O/sub 3/, pyrolytic carbon, and metals: surface coatings of siliconized silicon carbide, pyrolytic carbon, and metal carbides; properties of and the effects of irradiation on graphite spheres; the use of natural graphite in preparing a high-density matrix material; graphite fueling by thorium nitrate infiltration; subsurface metal and metal carbide coatings for graphite; and an in-pile loop program on themore » behavior of fission products in a recycle helium stream. (auth)« less

Authors:
Publication Date:
Research Org.:
Sanderson and Porter, New York, NY (United States)
Sponsoring Org.:
US Atomic Energy Commission (AEC)
OSTI Identifier:
4804887
Report Number(s):
NYO-9064
NSA Number:
NSA-16-009663
DOE Contract Number:
AT(30-1)-2378
Resource Type:
Technical Report
Resource Relation:
Other Information: Changed from OFFICIAL USE ONLY Aug. 18, 1961. Orig. Receipt Date: 31-DEC-62
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; ALUMINUM; ALUMINUM OXIDES; BURNUP; CARBIDES; CARBON; CERAMICS; COATING; CONFIGURATION; CONTAMINATION; COOLANT LOOPS; DENSITY; DIFFUSION; DISPERSIONS; FABRICATION; FISSION PRODUCTS; FISSIONABLE MATERIALS; FUEL ELEMENTS; FUELS; GAS COOLANT; GAS FLOW; GRAPHITE; GRAPHITE MODERATOR; HELIUM; IMPURITIES; IN PILE LOOPS; INERT GASES; IRRADIATION; LEAKS; MATERIALS TESTING; MECHANICAL PROPERTIES; METALS; MOLECULES; PEBBLE BED; PELLETS; POWER PLANTS; PYROLYSIS; RADIATION EFFECTS; REACTOR CORE; REACTOR SAFETY; SILICON CARBIDES; SPHERES; SURFACES; THERMAL STRESSES; THICKNESS; THORIUM NITRATES; URANIUM 235

Citation Formats

None, None. Fuel element development program for the Pebble Bed Reactor. Final Report. United States: N. p., 1961. Web. doi:10.2172/4804887.
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None, None. Fuel element development program for the Pebble Bed Reactor. Final Report. United States. doi:10.2172/4804887.
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None, None. Sun . "Fuel element development program for the Pebble Bed Reactor. Final Report". United States. doi:10.2172/4804887. https://www.osti.gov/servlets/purl/4804887.
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@article{osti_4804887,
title = {Fuel element development program for the Pebble Bed Reactor. Final Report},
author = {None, None},
abstractNote = {The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found that molecularly deposited ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces and where a larger thickness-todiameter ratio could be used than if the coating were on the surface of the graphite sphere. Fuel elements were irradiated to burnups ranging up to about 6 at.% U/sup 235/. In all specimens containing a uniform dispersion of fuel, the graphite spheres were found to retain their structural properties after irradiation. Data are given on fuel particle coatings of A1/sub 2/O/sub 3/, pyrolytic carbon, and metals: surface coatings of siliconized silicon carbide, pyrolytic carbon, and metal carbides; properties of and the effects of irradiation on graphite spheres; the use of natural graphite in preparing a high-density matrix material; graphite fueling by thorium nitrate infiltration; subsurface metal and metal carbide coatings for graphite; and an in-pile loop program on the behavior of fission products in a recycle helium stream. (auth)},
doi = {10.2172/4804887},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sun Apr 30 00:00:00 EST 1961},
month = {Sun Apr 30 00:00:00 EST 1961}
}
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Technical Report:
View Technical Report (40.93 MB)
DOI:  10.2172/4804887
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  • Recent fuel element irradiations demonstrated the structural integrity of spherical uranium-graphite fuel elements at burn-ups in excess of the design requirements of a 125 Mw(e) Pebble Bed Reactor (PBR) power plant. Other irradiations indicated the successful development of a coated fuel particle which permits fabrication of fuel elements meeting the PBR design objectives of a fission-product release rate (R/B) ? /sup 10-6/. beta plus gamma system activity in a 125 Mw(e) PBR is 490 c, assuming complete release of /sup lO-6/ of all fission products volatile at or below 2500 deg F. The low R/B being obtained from PBR fuelmore » elements indicates that decay during diffusion of the short-lived volatile precursors of nonvolatile daughter products will result in further reduction of this system activity, and an increase in the average half life of the fission products remaining in the system will increase the efficiency of a bypass clean-up system. The method of fabricating coated particles by the hydrolysis of metallic chlorides to produce oxides or pyrolysis of hydrocarbons to produce carbon on a suitable substrate, is described, and preirradiation test results are given. An accelerated coated-fuel-particle program is discussed as well as development work on the Pebble Bed REactor concept as a whole. auth« less

  • ; ; ; ...
    An in-pile loop using low-pressure helium was designed and operated in the Brookhaven Graphite Reactor to carry out tests of two Pebble Bed Reactor fuel spheres containing UO/sub 2/ shot admixed with graphite. The fuel particles in one sphere were coated with Al/sub 2/O/sub 3/, while those in the other sphere were left uncoated. Data were obtained for the release rates of fission products from both spheres; the results are consistent with these expected for fission recoils. (D.L.C.)

  • Fabrication of alumina-coated UO/sub 2/ and pyrolytic carboncoated UC/ sub 2/ particles was studied. Some reaction was noted between alumina and graphite at 2500 deg F. For UC/sub 2/ particles coated with carbon at 2000 deg F, the coatings were found to crack at temperaturss above 2000 deg F, whereas 2450 deg F deposition gave fewer failures at 3600 deg F, more rapid deposition- ratss, and absence of excess soot formation. A Pebble Bed Reactor fuel element consisting of a 1.5-in. graphite sphere fueled with alumina-coated UO/sub 2/ particles was irradiated at 1400 deg F to a burrup of 3.3more » at.% U/sup 235/. Up to 2.5 at.% burnup, the fission product leakage factors (rate of release/rate of production) ranged between 10/sup -9/ and 10/sup -5/ for 10 isotopes and were ascribed to trace amounts of uranium contamination outside the particle coatings. In the latter part of the quarter, ths leakage factor for Xe/sup 133/ rose to lO/ sup -3/ while several other shorter lived fission products increased to a smaller extent, indicating the beginning of diffusion through the coatings due to radiation damage of the alumina. Carbon-coated UC/sub 2/ particles coated at 2000 deg F were found to have good fission product retention up to 2000 deg F, and less than 4.5 x 10/sup -6/ of the Xen/sup 133/ was release d in a 4-hr test. Graphite bodies of density>2.0 g/cm/sup 3/ were prepared using natural graphite powder as filler material, and their crushing strength was on the same order as that of AGOT synthetic graphite. (D. L. C.)« less

  • Pyroltic Carbon Coated UC/sub 2/ Particles. Two types of commercially prepared pyrolytic carbon coated UC/sub 2/ particles were evaluated. They are identified as Batch PyC-7 and Batch PaC-6. The Batch PyC-6 particles were seen to have bumpy surfaces. Metallographic examination under polarized light showed a conical carbon growth structure typical of coatings deposited at about 3600 deg F. The Batch PyC-7 particles were seen to be smooth and the conical growth structure was not evident under polarized light indicating a coating deposited below 2900 deg F. Alpha assays of ths particles showed surface uranium contaminations of 10/sup -5/ to 10/supmore » -4/ of ths uranium contained in each particle. The values are somewhat higher than the typical values found for Al/ sub 2/O/sub 3/ coated UO/sub 2/. Acid leaching indicated that in some cases the uranium contamination was within the coating surface and could not be reduced by leaching. In other cases, the leach test indicated that a number of faulty particles were present since a large increase was noted in uranium contamination after leaching because of uranium in solution being deposited back on the coating surfaces. Fueled Spberes. A graphite sphere fueled with PyC/UC/sub 2/ particles (Type FA-25) was examined. The coated particles were uniformly dispersed throughout the FA-25 spheres. There was no evidence of coated particle damage in the region where the molding flash was removed. The FA-25 spheres were designed to have a subsequent protective coating of siliconized silicon carbide. A neutron activation test of one specimen showsd a Xe-133 release fraction of 1.1 x 10/sup -3/ on heating for 3 hr at 1650, 2100, and 2400 deg F. Similar data on spheres fueled with AliO/sub 3/ coated UO/sub 2/ showed Xe-133 release fractions of about 10/sup -6/ in a 4 hr heating peridod. Subsequent evaluation of the FA- 25 specimens produced no clear evidence that the PyC/UC/sub 2/ particles were damaged in fabrication and appeared to indicate that there were faulty particle coatings prior to sphere fabrication. Al/sub 2/O/sub 3/ coated UO/sub 2/. A commercially prepared Al/sub 2/O/sub 3/ coated UO/sub 2/ was exmanied. The coating was applied by the vapor deposition technique at 1830DEF. A rotating kiln was used in coating ths particles rather than a fluidized bed as in previous Al/sub 2/O/sub 3//UO/sub 2/ work. The 105- to 149-Mcron UO/sub 2/ shot was coated with 40 microns of Al/sub 2/O/sub 3/ in eight steps. A 5-gin sample experienced only a 0.0024-gm weight gain after a 5-hr exposure to l2O0 deg air. There was no detectable surface uranium contamination on the as-received particles using the alpha assay technique. The total exposed uranium, i.s. residual surface contamination plus uranium in solution, after leaching in hot nitric acid was 10/sup -5/ of the contained uranium. In-Pile Loop. The installation of an In-Pile Loop in the Brookhaven Graphite Reactor was completed. The Loop was designed to study the behavior of fission products released from a PBR fuel element into a recycled helium stream. Irradiation was started using an unfueled electricallly heated, graphite sphere in place of a fueled specimen to complete shakedown testing prior to fueled-sphere operation. Operations were curtailed when it was discovered that a leak had developed in the vacuum insulation annulus of the in-pile section which appeared to have occurred during the exceptionally difficult insertion prior to startup. The effect of the leak was to limit operating temperatures to about 500 deg F gas outlet and 800 deg F in the specimen, instead of the design values of 1250 deg F and 1800 deg F. The leak could not be repaired because the in-pile section was too highly activated. Consequently, a new in-pile section was fabricated and, concurrently, a graphite sphere fueled with Al/sub 2/)/sub 3/ coated UO/sub 2/ was irradiated in the defective in-pile section.« less