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Design Criteria for a Fast Flux Liquid Metal Loop in the Advanced Test Reactor

Description: Design criteria are presented for liquid-metal loop facilities for the Advanced Test Reactor. It was concluded that achievement of the design objectives could not be predicted with any high degree of confidence when utilizing the package loop concept based on the design philosophy of the PW-19 loop. (M.C.G.)
Date: July 1, 1962
Partner: UNT Libraries Government Documents Department

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT

Description: General research and development are reported on watercooled and sodium- cooled reactors. Studies are also reported on reactor safety and nuclear technology. (For preceding period, see Report ANL-6253.) (W.L.H.)
Date: November 1, 1960
Creator: Hilberry, N.
Partner: UNT Libraries Government Documents Department

Summary Report of Design Criteria for a Thermal Flux Liquid Metal Package Loop in the Advanced Test Reactor

Description: Results of the design criteria work performed prior to the termination of work on the package concept for the liquid metal loop facilities for the Advanced Test Reactor are summarized. Based on the results of the criteria studies, it was concluded that achievement of the design objectives could not be predicted with any high degree of confidence when utilizing the package loop concept based on the design philosophy of the PW-19 loop. The characteristics of the loop are given along with design and development accomplishments for associated equipment. (N.W.R.)
Date: March 1, 1963
Partner: UNT Libraries Government Documents Department

Design Criteria for a Fast Flux Liquid Metal Loop in the Advanced Test Reactor

Description: Criteria are presented for the Titles I and II design of the Fast Flux Loop Facility which is to be a part of the ATR. This facility, a NaK-cooled in- reactor packaged loop based on the Pratt & Whitney Aircraft PW-19 loop design, is to be used for irradiation of fast reactor assemblies under simulated fast reactor operating conditions. (D.L.C.)
Date: July 1, 1962
Partner: UNT Libraries Government Documents Department

Ten-Year Sodium-Reactor Development Program

Description: >A 10-year program of development and construction of large-scale, sodium-cooled reactors is summarized. The current state of development of the SGR and its associated components is sufficiently advanced to permit construction of economic plants within the 10-year period. Two advanced Sodium Reactor concepts are presented. A construction program involving two reactor experiments and two full-scale plants with a capacity of 550 Mwe, together with associated development, is estimated to cost 6 million. Of this amount approximately 06 million would be borne by the AEC and the remainder by power utility companies. Escalation and construction loan interest charges are included in these figures. The cost of power from the larger power plant would be approximately 6 mills/kw-hr, based on 1959 dollars. (auth)
Date: April 11, 1959
Partner: UNT Libraries Government Documents Department

STEAM GENERATOR PRELIMINARY DESIGN

Description: A conceptual study on design of sodium-cooled reactor steam generators was conducted. Included is a detailed description of the preliminary design and analysis, based on the use of known materials and existing methods of fabrication. (See also APAE-41 Vols. I and III.) (J.R.D.)
Date: February 28, 1959
Partner: UNT Libraries Government Documents Department

AN EVALUATION OF MERCURY COOLED BREEDER REACTORS

Description: Under the New Reactor Concepts Evaluation Program sponsored by the United States Atomic Energy Commission. Advanced Technology Laboratories (a Division of American Radiator & Standard Sanitary Corporation) has undertaken am investigation of the technical feasibility and economic potential of the use of boiling mercury as a coolant for fast breeder reactors The investigation was performed between January 1, 1959, and October 31. 1959. This is the final report on that investigation and is submitted in compliance with the terms of the program authorization, Contract Number AT(04-3)-109, Project Agreement Number 4. (auth)
Date: October 31, 1959
Creator: Bradfute, J.O.; Battles, D.W.; Clark, G.S.; Corridan, R.E.; Gellenbeck, E.T.; Kavanagh, D.L. et al.
Partner: UNT Libraries Government Documents Department

NaK FREE CONVECTION COOLED SHAFT FREEZE SEAL FOR SRE PUMPS

Description: An investigation was conducted to determine the feasibility of using an NaK free-convection-cooled shaft freeze seal on the SRE main sodium pumps. The use of sodium in this application instead of tetralin eliminates the carburization hazard present should the seal coolant-(tetralin) leak into the sodium system. Results and recommendations are included. (J.R.D.)
Date: June 18, 1959
Creator: Perez, F.
Partner: UNT Libraries Government Documents Department