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METALLURGY DIVISION ANNUAL REPORT FOR 1959

Description: Engineering Metallurgy. Investigations were made on the following: development programs for Experimental Breeder Reactor No. 2 (EBR-II); fabrication of care and blanket components for Argonne Fast Source Reactor (AFSR); fabrication of spiked fuel rods for Experimental Boiling Water Reactor (EBWR) (Core lA); fuel for Transient Test Reactor Facility (TREAT); fuel elements for Experimental Breeder Reactor No. 1 (EBR-I) (Mark-IV); canning of plutonium loading for Zero Power Reactor No. 3 (ZPR-III); development of ceramic fuel materials; irradiation evaluations of various experimental fuel and control-rod materials: postirradiation examinations of full-scale reactor fuel elements: development of corrosion-resistant fuel and jacketing materials; nondestructive testing developments; and metallurgical assistance to the Fast Reactor Safety Program. Basic Metallurgy. Investigations were made on the following: preparation of high-purity materials; physical metallurgy of uranium; constitution and properties of uranium and plutonium alloys; alloying properties; x-ray and neutron-diffraction studies; problems in metal physics; corrosion research; irradiation effects; and ceramicmaterials research. (W.L.H.)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department

COMMENTS ON THE ANGLE OF REPOSE OF AQUEOUS ThO$sub 2$ SLURRIES

Description: aqueous thorium oxide slurries have arisen in connection with the design of the lower portion of the blanket pressure vessel and its associated piping. The literaiure review on the angle of repose of various dry solids by Mrs. B. S. Neumann in J. J. Hermans' book on''Flow Properties of Dispersed Systems'' is summarized. Mrs. Neumann points out that both a dynamic and static angle of repose have been observed with the angles for dynamic systems being larger than for static systems. Results are reported for two different types of experiments with slurry discharged from run 200A-14 which gave static angles of repose of 45 to 50 deg and dynamic angles of 65 to 75 deg . (auth)
Date: February 14, 1958
Creator: Thomas, D.G.
Partner: UNT Libraries Government Documents Department

U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

Description: Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by ...
Date: January 1, 1995
Partner: UNT Libraries Government Documents Department

Design windows for a He cooled fusion reactor

Description: A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study of a parameter regime under which a possible design exists with different design requirements, such as allowable pumping fraction. The concept identifies not only the required parameter regime, but also investigates the robustness of the design, i.e., the validity of the design with change of design parameters and requirements. Some recent directions of helium cooled design for ITER and for divertor can also be explained by this design window concept.
Date: October 1, 1993
Creator: Sze, Dai-Kai & Hassanein, A.
Partner: UNT Libraries Government Documents Department

BEATRIX-II: In-situ tritium recovery data correction

Description: BEATRIX-II was an in-situ tritium recovery experiment in a fast reactor to characterize the irradiation behavior of fusion ceramic breeder materials. Correcting and compiling the in-situ tritium recovery data involved correcting the ion chamber response for the effect of sweep gas composition or amount of hydrogen in the helium sweep gas and for the buildup of background. The effect of sweep gas composition was addressed in the previous workshop. During the operation of Phase I of the experiment the backgrounds of the ion chambers were found to reach significant levels relative to the tritium recovery concentrations in the sweep gas from the specimen canisters. The measured tritium concentrations were corrected for background by comparing the tritium recovery rate during reference conditions with the predicted tritium generation rate. Background increases were found to be associated with tritium recovery peaks and elevated levels of moisture in the sweep gas. These conditions typically occurred when the hydrogen concentration in the sweep gas was increased to 0.1% after extended operation in He or He-0.01% H{sub 2}. Three examples of this increase in ionization chamber background are described. The final corrected BEATRIX-II, Phase I tritium recovery data provide a valuable resource to be used for predicting the performance of Li{sub 2}O in a fusion blanket application.
Date: September 1, 1993
Creator: Slagle, O. D.; Hollenberg, G. W.; Kurasawa, T. & Verrall, R. A.
Partner: UNT Libraries Government Documents Department

Studies of fusion reactor blankets with minimum radioactive inventory and with tritium breeding in solid lithium compounds: a preliminary report

Description: The feasibility of fusion reactor blankets with low residual activity is examined. Several designs are examined with regard to activation, tritium breeding ratio, mechanical design, tritium removal from the blanket, and thermal cycle efflciency. Using aluminum (SAP) as a structural material, it should be possible to build CTR blankets with ~10 curies/MW(e) of long-lived (half life one day or greater) residual activity (other than tritium), which is many orders of magnitude less than with Nb or stainless steel blankets. In the designs examined in this study, tritium is bred in -solid-lithium-containing materials, e.g., LiAl alloy, which have high-equilibrium tritium pressures. The tritium diffuses into either the helium-coolant stream, from which it is removed by absorption, or into the vacuum region between the first wall and plasma. Depending on processing methods and blanket parameters, the tritium blanket inventory should range from 10/sup 2/ to 10/sup 3/ curies/ MW(e). Tritium breeding ratios range from 0.9 to 1.5 depending on blanket design, while therrnal cycle efficiency is estimated to range from 35% to well over 40%, depending on design. Several module designs are developed in which helium-coolant exit temperatures are substantially above the operating temperature limit (~400 deg C) for the aluminum (SAP) structure. (auth)
Date: June 1, 1973
Creator: Powell, J.R.; Miles, F.T.; Aronson, A. & Winsche, W.E.
Partner: UNT Libraries Government Documents Department

An assessment of the base blanket for ITER

Description: Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored.
Date: December 31, 1991
Creator: Raffray, A. R.; Abdou, M. A. & Ying, A.
Partner: UNT Libraries Government Documents Department