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Three separate irradiation experiments were completed with Th and Th-U alloys. In the first experiment, three-rolled plates of Th and Th-5 wt% U alloy irradiated to total atom burnups up to 1.5% at 200 deg C showed no anisotropic growth and decreased in density at a rate of 1% per wt.% burnup. In the second experiment, 15 swaged specimens of Th and of the alloys Th-0.1 wt% U, Th-1.4 wt% U, and Th-5.5 wt% U were irradiated to burnups ranging from 0.3 to 3.6% of all atoms at temperatures in the range of 45 to 200 deg C. Again, no anisotropic growth was observed and densities decreased at rates near 1% per wt.% burnup. A Th-1.4 wt% U alloy specimen with 2.0 wt.% burnup was found to have retained significant room-temperature ductility. In the final experiment, a group of 44 chill-cast specimens of Th alloys containing 10, 15, 20, 25, and 31 wt% U were irradiated to burnups ranging from 0.16 to 10.1% of all atoms. Maximum irradiation temperatures ranged from 260 to over 1000 deg C. Surface roughening occurred in the alloys containing 25 and 31 wt% U. Volume increases at any given temperature for all alloys were linear with increasing burnup. The rate of volume increase for all alloys rose from approximately 1% per wt.% burnup at the lower temperatures to a value of 2.5 at 650 deg C. Thereafter the swelling rate increased somewhat, reaching a value of 6% volume increase per wt.% burnup at 800 deg C. The rates of volume increase under irradiation of Th-U alloys in the entire temperature range studied were significantly less than those reported for the best U-base alloys. It is suggested that the excellent resistance to high- temperature swelling of the cast Th-U alloys resulted from the fact that a dispersion of very thin U particles was obtained. A high probability, therefore, existed for fission recoils to escape from the U particles into the isotropic and less densely packed Th matrix.
The swelling of uranium and of a few selected uranium alloys on post-irradiation annealing was investigated by utilizing density measurements in conjunction with the observation of pores in the microstructures of annealed specimens. Specimens were irradiated to about 0.3 at.% burnup in a constrained condition at approximately 275 deg C and were subsequently pulse annealed. The amount of swelling was found to be less than 1% for U specimens that were pulse annealed up to 75 hr at temperatures below 550 deg C; the amount of swelling, however, increased considerably on annealing at temperatures between 550 and 650 deg C. Specimens pulse annealed up to 75 hr at 618 deg C decreased in density by approximately 18%. The swelling was accompanied by the formation of bubbles on grain boundaries in recrystallized regions. The observations suggest that recrystallization is a necessary prerequisite for pronounced swelling in the alpha phase.
The Metallurgy of Nuclear Fuel: Properties and Principles of the Technology of Uranium, Thorium and Plutonium is a systematic analysis of the metallurgy of nuclear fuel, with emphasis on the physical, mechanical, and chemical properties as well as the technology of uranium, thorium, and plutonium, together with their alloys and compounds. The minerals and raw material sources of nuclear fuel are discussed, along with the principles of the technology of the raw material processing and the production of the principal compounds, and of the pure metals and alloys. Comprised of three parts, this volume begins with an introduction to the history of the discovery of uranium and its position in the periodic system; its use as a nuclear fuel; radioactivity and isotopic composition; alloys and compounds; and physical, mechanical, and chemical properties. The effect of mechanical and thermal treatment, thermal cycling and irradiation on the physicochemical properties of uranium is also examined. The next two sections are devoted to thorium and plutonium and includes chapters dealing with their uses, alloys and compounds, and methods of recovery and purification. This book is written for university students, but should also prove useful to young production engineers and scientific workers who are concerned with problems in the metallurgy of nuclear fuel.