Author: Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 68
Book Description
Annual Progress Report on Fuel Element Development for FY ...
Author: Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 68
Book Description
Annual Progress Report on Fuel Element Development for FY 1963
Author: G. W. GibsonAnnual Progress Report on Fuel Element Development for FY 1962
Author: G. W. GibsonANNUAL PROGRESS REPORT ON FUEL ELEMENT DEVELOPMENT FOR FISCAL YEAR 1961
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Progress in fuels and materials development is summarized. Major areas of investigation include a materials study by means of sample fuel plates containing uranium alloys or cermets, burnable poisons, non-uniform fuel and poison distributions and clad with various aluminum alloys; and an engineering study of fuel element geometries optimized in heat transfer, hydraulics, and materials strength. Up to 45 wt% U-Al alloys, 6 to 65 wt% UO/-Al and U3O6-Al dispersions, including enrichments ranging from 20% to 93%, were tested to 70% burnup in de-ionized water at 200 deg F in the MTR. Their performance at higher temperature is still being investigated. Test results for the MTR conditions indicate that all of the compositions investigated to date will successfully withstand even the longest irradiation at these conditions if properly fabricated. Some high strength aluminum alloy claddings, not yet fully tested, show some peculiar surface effects which may be related to corrosion. Metallographic studies of irradiated cermets reveal a reaction'' (diffusion) zone produced around UO2 particles in contact with aluminum. These zones are being studied by means of x-ray diffraction, electron microscopy, and electron microprobe analysis. From engineering studies has come promise of improved heat removal and lower pumping requlrements for reactors through artificial roughening of fuel plates. Computer optimizatlon studies and hydraulic tests indicated 80% improvement in heat transfer or 60% less flow for the same heat load are obtainable for MTR conditions. Heat transfer test results from 0.110 x 2.624 ' electrically-heated channels using heat fluxes up to 2.88 x 106 Btu/hr-ft/ sup 2/, sgree better with correlations based on bulk temperatures than with the more widely used modified Colburn equation. In this range, a modifled Colburn equation with a 20% safety factor, as is presently used, seems adequate. However, an equation based on the bulk coolant temperature could be used employing a smaller safety factor because of its greater accuracy. (auth).
Annual Progress Report on Fuel Element Development for FY 1961
Author: G. W. GibsonANNUAL PROGRESS REPORT ON FUEL ELEMENT DEVELOPMENT FOR FY 1963
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Progress in fuels and materials development is reported. Irradiation tests on powdered UAl3 intermetallic compounds demonstrated good stability and fission gas retention capabilities. Developmental aluminum powder metal products showed good corrosion resistance at high temperatures while retaining excellent high temperature strength. All of the fuel compositions tested (UO/sub 2/, U3O, and UAl3 in aluminum matrices) exhibited density decreases under irradiation. Tensile tests on sandwich-type fuel plates at elevated temperatures indicated that the fuel plate strength is strongly influenced by the core material rather than dependent primarily on the cladding material as was found true of lower (MTR) temperatures. Three capsules containing beryllium were inserted in the ETR, in order to determine strength, gas release, and growth during a high-temperature (600--800 deg C) irradiation. An MTR fuel element employing advanced metallurgical techniques to optimize the hydraulic and heat transfer characteristics was fuily irradiated in the MTR. The fuel element consisted of 32 plates containing 250 g U235 in a U3O/ sub 8/--Al dispersion. A prototype ETR fuel element was made without side plates. (M.C.G.).
Annual Progress Report on Reactor Fuels and Materials Development for ...
Author: Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 58
Book Description
ANNUAL PROGRESS REPORT ON FUEL ELEMENT DEVELOPMENT FOR FY 1962
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Activities in a project aimed at the improvement of fuel elements for high flux test reactors are reported. The investigation of new fuel compositions, distributions, and geometries is being undertaken to increase fuel life, to improve the flux distribution, and to provide a means of safely reaching higher reactor operating power and power density in these reactors. The effects of nuclear irradiation on the fuel and structural materials is being studied to predict the performance of these materials in more advanced reactor designs. A summary of the past year's progress is given and the fabrication and irradiation of samples containing up to 50 wt % U--Al alloys, cermets of UO2, U3O/ sub 8/, UC, UN, U3Si, and Al, clad in various Al an d Be--Al materials is described. The use of ThO2 and Th cores, the addition of BeO to cermet cores and high density fuei cores of U--Al intermetailics produced by powder metallurgy techniques were studied during the year. High strength APM claddings involving Al2O3 contents from 8 to 10% were tested and indicate the need for improved quality control of the APM material. Duplex claddings involving burnable poison layers and APM clad with corrosion resistant X8001 showed promise where special properties are desired. The results of the work continue to demonstrate the excellent radiation stability of U--Al fuels even after long irradiation exposure at elevated temperatures. Tests up to 350 deg F and after 50% burnup of the U235 in U--Al alloys, show no appreciable dimensional or microstructure changes. UO2 and U3O react with Al under radiation to form UAl4. Tensile tests of these fuels at ambient temperatures show appreciable loss in ductility with irradiation; several compositions actually exhibiting zero ductility. Irradiation at temperatures up to 200 deg F of cold-worked and of heat-treated Al does not destroy the pre- irradiation hardness and strength of these materials. Computer optimization of fuel element geometries from the standpoint of heat transfer, hydraulics, and strength resulted in the design and fabrication of a 32-plate fuel element. Hydraulic tests produced favorable results and the element is ready for MTR testing. Future work stressing materials development will be directed toward extending U--Al fuels to use at 400 to 800 deg F. Continued studies on graded fuels, Be damage, and the Th--U233 system are also planned. Tensile testing will be extended to higher temperatures in pre- and post-irradiation measurements and the study of the effect on cold worked and tempered materials of elevated temperature-radiation exposures will be continued. (auth).
Summaries of Fuels and Materials Development Program
Author: William L. R. RicePublisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 286
Book Description
Summaries of Fuels and Materials Development Programs
Author: William L. R. RicePublisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 278
Book Description