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TCT hybrid preconceptual blanket design studies

Description: The conceptual design of a tokamak fusion-fission (hybrid) reactor, which produces electric power and fissile material, has been performed in a cooperative effort between Princeton's Plasma Physics Laboratory (PPPL) and Battelle's Pacific Northwest Laboratories (PNL). PPPL, who had overall project lead responsibility, designed the fusion driver system. Its core consists of a tokamak plasma maintained in the two-component torus (TCT) mode by both D and T beams and having a single null poloidal d… more
Date: January 1, 1978
Creator: Aase, D. T.; Bampton, M. C. C.; Doherty, T. J.; Leonard, B. R.; McCann, R. A.; Newman, D. F. et al.
Partner: UNT Libraries Government Documents Department
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Repair Welding of Fusion Reactor Components

Description: Recent experimental investigations indicate that the repair welding of irradiated materials containing greater than 1 to 2.5 appm helium leads to catastrophic cracking in the heat affected zone of the weld. The high temperatures and cooling tensile stresses which occur during the welding process lead to enhanced helium bubble growth in the heat affected zone region, resulting in catastrophic cracking upon cooling. An investigation is proposed which seeks to determine the effect of stress state … more
Date: May 20, 1992
Creator: Chin, Bryan A.
Partner: UNT Libraries Government Documents Department
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Investigation of joining techniques for advanced austenitic alloys

Description: Modified Alloys 316 and 800H, designed for high temperature service, have been developed at Oak Ridge National Laboratory. Assessment of the weldability of the advanced austenitic alloys has been conducted at the University of Tennessee. Four aspects of weldability of the advanced austenitic alloys were included in the investigation.
Date: May 1, 1991
Creator: Lundin, C. D.; Qiao, C. Y. P.; Kikuchi, Y.; Shi, C. & Gill, T. P. S.
Partner: UNT Libraries Government Documents Department
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Electrolytic Dissolution of Power Reactor Fuels in Nitric Acid

Description: The electrolytic oxidation in nitric acid of stainless steel, zirconium, Zircaloy-2, zirconium- uranium alloy, aluminum, and uranium - molybdenum alloy was demonstrated on a laboratory scale. The rate of chemical dissolution of UO/ sub 2/ in nitric acid was measured. Corrosion of stainless steel by these dissolver solutions was measured and found to be negligible. Electrolytic dissolution was demonstrated to be a practical technique for the first step in processing fuel elements of several type… more
Date: October 1, 1961
Creator: Clark, A. T., Jr.; Meyer, L. H.; Owen, J. H. & Rust, F. G.
Partner: UNT Libraries Government Documents Department
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Corrosion Behavior of Reactor Materials in Fluoride Salt Mixtures

Description: Molten fluoride salts, because of their radiation stability and ability to contain both Th and U, offer important advantages as high-temperature fuel solutions for nuclear reactors and as media suitable for nuclear fuel processing. Both applications have stimulated experimental and theoretical studies of the corrosion processes by which molten salt mixtures attack potential reactor materials. Corrosion experiments with fluoride salts which were conducted in support of the Molten-Salt Reactor E … more
Date: September 19, 1962
Creator: DeVan, J. H. & Evans, R. B., III
Partner: UNT Libraries Government Documents Department
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Delayed Failure Hydrogen Embrittlement of Zirconium. Summary Report, September 15, 1961 to September 14, 1962

Description: The extent to which zirconium and zirconium alloys exhibit delayed failure (static fatigue) as caused by a combination of absorbed hydrogen and applied stress was investigated. Susceptibility to time-dependent fracture was evaluated for unalloyed zirconium and Zircaloy-2 with 200 and 500 ppm hydrogen as well as for an experimental Zr Al-Sn-Mo alloy and the Canadian Zr-2.5Nb cladding material. For unalloyed zirconium and Zircaloy-2 containing up to 500 ppm hydrogen, no room-temperature, timedepe… more
Date: October 10, 1962
Creator: Weinstein, D. & Holtz, F. C.
Partner: UNT Libraries Government Documents Department
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CASTING OF LONG AND THIN PLATES OF URANIUM-MOLYBDENUM ALLOYS

Description: The development of procedures for the vacuum induction casting of U--Mo alloys into both thin (0.010 to 0.100-in. thick) plates and long (36 in.) plates is described. Melting and casting cycles were developed, and casting techniques established, which resulted in sound, integral plates. These plates were evaluated by radiographic and metallographic examination, and by chemical analysis. The results indicated the feasibility of the process for the fabrication of fuel plates for nuclear reactors.… more
Date: November 1, 1961
Creator: Katz, N.H. & Binstock, M.H.
Partner: UNT Libraries Government Documents Department
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REACTIVITY CALIBRATIONS AND FISSION-RATE DISTRIBUTIONS IN AN UNMODERATED, UNREFLECTED URANIUM-MOLYBDENUM ALLOY RESEARCH PROGRAM

Description: Completion of zero-power critical experiments with the ORNL Health Physics Research Reactor is reported. A description is given concerning these experiments which were used to determine the critical size, fission-rate distributions, reactivity calibrations of its movable parts, the temperature coefficient of reactivity, and the reactivity effects of the presence of neutron- reflecting materials adjacent to the reactor. (J.R.D.)
Date: May 10, 1962
Creator: Mihalczo, J.T.
Partner: UNT Libraries Government Documents Department
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FAST BREEDER REACTOR THERMOCOUPLE DEVELOPMENT.

Description: Studies of very high-temperature thermocouple characteristics for instrumentation and electrical components used in fast breeder reactors showed that the thermoelectric emf produced along BeO, HfO{sub 2}, or ThO{sub 2} insulators can affect the output of the thermoelements. Bare-wire thermocouples are influenced by environment. Radiation effects on W versus W - Re thermocouples in the ORR after ~ 2 months in a 1.2 x 10{sup 14}n/cm{sup 2}-sec thermal and 2.1 x 10{sup 13}n/cm{sup 2}-sec fast flux… more
Date: October 31, 1969
Creator: Funston, E.S. & Kuhlman, W.C.
Partner: UNT Libraries Government Documents Department
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Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

Description: Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical prope… more
Date: June 25, 1980
Partner: UNT Libraries Government Documents Department
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Cobalt-60 heat source demonstration program. Phase III. Fabrication. Final report

Description: Significant accomplishments completed during Phase III of the /sup 60/Co Heat Source Demonstration program include the following: encapsulation of 2 MCi of /sup 60/Co; fabrication of the heat source, including the ASME coded pressure vessel/core assembly, and biological shielding; endurance testing of a prototype heat pipe for a period of 28 months; fabrication and qualification of the heat pipe emergency cooling subsystem; issue of the safety evaluation report, reference 3, and the operations … more
Date: June 1, 1973
Partner: UNT Libraries Government Documents Department
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ADVANCED PRESSURE VESSEL MATERIALS.

Description: Data on 5Cr - 3Mo - 12Ni alloy, PH13 - 8Mo and Iconel 718 for pressure vessel application are given. The 5Cr - 3Mo - 12Ni posseses characteristics suited to uniform aging of heavy sections. Tensile and toughness properties are good, and structural stability up to 315 deg C is adequate. Weldability with 12Ni - 3Cr - 3Mo filler material is good but 12Ni - 5Cr - 3Mo filler is more efficient. Aging characteristics of PH13 - 8Mo may not be suited to uniform aging of heavy sections. Excellent tensile… more
Date: October 31, 1969
Creator: Robertshaw, F.C.; Stephan, H.R. & McConnelee, J.E.
Partner: UNT Libraries Government Documents Department
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