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PLUTONIUM GRAPHITE ASSEMBLIES--PART II

Description: Neutron multiplication measurements were made on a number of cylindrical assemblies of Pu and graphite disks. S/sub n/ calculations were made on homogeneous mixtures of Pu and graphite with varying C/Pu ratios and varying reflector thickness. (auth)
Date: August 10, 1959
Creator: Goodwin, A. Jr. & Schuske, C.L.
Partner: UNT Libraries Government Documents Department

GRAPHITE-MODERATED, GRAPHITE-REFLECTED CRITICAL ASSEMBLIES

Description: Graphite-moderated, graphite reflected critical assemblies have been set up in the LASL Honeycomb remotely controlled machine. lnformation has been obtained on the critical masses of systems having C/Oy ratios of 6650 and 4093. A third system at a smaller ratio is planned. The reactivity contribution of channels through the core and reflector was determined. (auth)
Date: January 1, 1957
Creator: Byers, C.C.
Partner: UNT Libraries Government Documents Department

NUMERICAL SOLUTION OF REACTOR STRESS PROBLEMS

Description: Generalized computer codes were devised for solving stress problems of some complexity. These codes were applied to stress problsms relating to the graphite moderator elements in the Experimental Gas-cooled Reactor. The stress relief obtained by aubdividing the moderator elements was evaluated. The distontion and bending moments of the elements were also determined. (auth)
Date: February 28, 1961
Creator: Redmond, R.F.; Hulbert, L.E. & Clark, R.W.
Partner: UNT Libraries Government Documents Department

EGCR CORE STRUCTURAL ANALYSIS. THE EFFECTS OF FAST-NEUTRON IRRADIATION AND THE BOWING CHARACTERISTICS OF THE GRAPHITE COLUMNS

Description: An analysis of the EGCR core structure was made to determine the lateral deflections (bowing) of the graphite columns resulting from shrinkage caused by fast-neutron irradiation, the life expectancy of each column due to restraints imposed on the bowing, and the reaction forces induced in the supporting structures. Based on currently avallable data for EGCR type graphite shrinkage and assuming experimental loop operation, a maximum bowing potential of 3.61 in. was calculated for an interior column. It was found that strains equivalent to the rupture strains observed from tensile tests could be expected after 4 to 6 years of full-power operation. Over half of the columns will reach these strains before the 20-yr reactor design life is reached. (auth)
Date: April 14, 1961
Creator: Moore, S.E. & Shaw, W.A.
Partner: UNT Libraries Government Documents Department

High Temperature Radiation Induced Graphite Contraction

Description: Information concerning graphite contraction applicable to high- temperature, graphite-moderatored reactors is presented. The scope includes relevant data from all available sources, interpretation and extrapolations as can reasonably be made, and a discussion of the effects observed in terms of current radiation damage theory. Limits of accuracy and a discussion of experimental techniques are presented. (auth)
Date: February 1, 1959
Creator: Davidson, J. M.; Woodruff, E. M. & Yoshikawa, H. H.
Partner: UNT Libraries Government Documents Department

Report of the Objectives and Plans for the AEC's Civilian Power Gas Cooled Reactor Program

Description: Progress in the U. S. civilian power gas-cooled reactor program is discussed. Gas reactors having technical features of high conversion ratio, high temperature, high fuel burnup, and capability of construction in large sizes make them very attractive as potential producers of economic power in the very near term. The operation of Peach Bottom-HTGR and EGCR in late 1964 and 1965, respectively, will contribute to the successful exploitation of thermal gas- cooled reactors. Since the graphite fuel concept promises very low fuel cycle costs along with reactor coolant conditions that can exceed current practice, it was concluded that the concept provides a long term potential that promises some very exciting possibilities. (auth)
Date: June 1, 1963
Creator: Pahler, R. E.
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

GRAPHITE PROBLEMS IN AIR-COOLED REACTORS

Description: The problems of oxidation, radiation induced dimensional expansions, and accumulation of stored energy affect long term use of graphite in air-cooled reactors. Results show that frequent low temperature annealing eliminates these problems and may extend the lifetime of the graphite indefinitely. (N.W.R.)
Date: January 1, 1962
Creator: Schweitzer, D.G.
Partner: UNT Libraries Government Documents Department

EGCR Graphite Permeability Tests: Results of Forced Flow Experiments on Egcr Moderator-Grade Graphite

Description: Helium-permeability and porosity were determired at room temperature for specimens from a typical EGCR moderator-grade graphite block. Permeability, at a mean pressure of 2 atm, ranged from 26 to 200 (av. 86.5) millidarcys. Permeability data indicated that turbalent flow was never obtained with helium in these tests and that helium permeating the moderator graphite at EGCR operating conditions (taken to be: 600 deg C; DELTA P, 10 lb/in./sup 2/ per inch of graphite; mean P, 400 lb/in./sup 2/) was in the viscous flow region. Daroy's law and the reported constants are applicable for flow computations involving moderator graphite under these conditions. Porosity ranged from 20.6 to 29.4% (av. 23.8%), and there was no correlation between porosity and pemaesbility variations. The large variations encountered were believed to reflect the nonuniformity of the specimens, since duplicate determinations showed excellent agreement. Permeabilfty did not change appreciably with direction of flow and did not vary consistently with respect to the extrusion or any other axis. Preparation of the specimens did not appear to introduce appreciable surface effects. (auth)
Date: March 24, 1961
Creator: Ward, W. T. & Truitt, J.
Partner: UNT Libraries Government Documents Department

Preliminary Evaluation of Chlorine for Use as a Gas Cooled Reactor Safeguard

Description: An evaluation of chlorine in a high-temperature gascooled graphite- moderated reactor as a safeguard to control u runawav oxidation reaction was made. Experiments were performed which demonstrated the ability of a small amount of chlorine in an air stream to reduce the oxidation rate of graphite. Chlorine appeared to inhibit graphite oxidation by blocking active sites on the surface. On the basis of this mechanism, a rate law was derived which was consistent with observed behavior. (C.J.G.)
Date: March 1, 1960
Creator: Dahl, R. E.
Partner: UNT Libraries Government Documents Department

Studies of Improvement of Power Density in Orr Loops

Description: Using a simplified model, calculations were made of the possible effects of voids upon the power density in ORR loop experiments. It is concluded that the power density may be markedly increased if voids and channels are plugged with moderator material such as graphite or beryllium. (auth)
Date: April 11, 1960
Creator: Tobias, M. L. & Vondy, D. R.
Partner: UNT Libraries Government Documents Department

The H-4, H-5, and H-6 Irradiation Experiments: Irradiation of N-Reactor Graphite, Interim Report Number 1

Description: Report regarding an experimental program by Hanford Laboratories' in order "to determine the long-term irradiation behavior of the graphite used as the moderator in N-Reactor. The primary objectives of the program are to provide data for predictions of the distortion of the N-Reactor moderator and the stress conditions which could arise from the difference in the the rate and extent of contraction between the transverse and parallel orientations of the graphite bars" (p. 1).
Date: October 1964
Creator: Helm, J. W.
Partner: UNT Libraries Government Documents Department

THE OAK RIDGE RESEARCH REACTOR (ORR), THE LOW-INTENSITY TESTING REACTOR (LITR), AND THE OAK RIDGE GRAPHITE REACTOR (OGR) AS EXPERIMENT FACILITIES

Description: >Characteristics of the ORR, LITR, and OGR that experimenters have found to be important are listed. The results of a survey conducted among experimenters on the utility of the reactors for various types of experiments are discussed, and some changes which might be made to improve the utilization are listed. A brief outline, with references, of most of the experiments currently being performed is included. (auth)
Date: August 28, 1962
Creator: George, K.D.
Partner: UNT Libraries Government Documents Department

Sodium Graphite Reactor Quarterly Progress Report for July-September, 1954

Description: Reactivity calculations have been performed for the steady-state Pu feedback technique outlined in the previous progress report. A full-scale power plant study was initiated, based on sodium-graprite technology. A twin-core power plant is now considered to be the most promising configuration. Several design drawings are given of such a reactor, using slightly enriched U to produce Pu amd electrical power. The thermal neutron flux distribution in a cluster of 6 U rods was measured, and the results are compared with previous measurements for 7 rod cluster. The average thermal cycling of hollow U slug elements was begun. Results are given for 500 cycles between 100 and 500 deg C. A series of powder- compacted U alloys were thermal cycled between 200 and 700 deg C. Data on the transfur of radioactivity from Zr by Na has been obtained from a capsule of the first series of three miniharps. Fe, Al, and Cu were immersed in toluene end irradiated at 150 deg F in the MTR-Gamma canal. Toluene is being considered as a shield coolant for the SRE. The effect of 1-Mev electron irradlation on terphenyls was also studied. A venting tube arrangement has been designed for the Zr-canned graphite moderator. A number of thermal insulating brick amd fiber materials were sublected to liquid Na to study deterioration effects. The materials tested were JohnsManville Brick C-16 (Sil-O-Cel mortar), Superex Paste, and Eagle-Pitcher Mineral Wool. Encouraging results were obtained in an efiort to evaluate the effectiveness of Na decontamination by liquid ammonia. Pressure drop and flow characteristics of the latest design SRE fuel element have been completed. Design details of the 2-speed control rod drive assembly are given. Other aspects of the reactor control system, including design and component fabrication, are discussed. Gamma dose rates at the surface of the top shield ...
Date: December 1, 1954
Creator: Siegel, S. & Inman, G. M.
Partner: UNT Libraries Government Documents Department

PLUTONIUM GRAPHITE ASSEMBLIES

Description: Neutron multiplication measurements and theoretical calculations were made on cylindrical assemblies of graphite and plutonium disks. (auth)
Date: September 29, 1958
Creator: Goodwin, A. Jr. & Schuske, C.L.
Partner: UNT Libraries Government Documents Department

Civilian Power Reactor Program. Part 3. Book 6. Status Report on Sodium Graphite Reactors as of 1959

Description: The current development status of the sodium graphite reactor concept is described. The development history is summarized, and all important areas of development are discussed. The discussion of the SGR program is broken into three categories: (1) general research and development, dealing with reactor physics, fuels and materials, components, etc., (2) experimental reactors, operating or in some phase of design or constnuction, and (3) power demonstration reactors, operating or in a design or constnuction phase. (W.D.M.)
Date: January 1, 1960
Partner: UNT Libraries Government Documents Department

Comparison of BEO Versus Graphite as a Moderator for MGCR

Description: The comparison is made on the basis of nuclear requirements, properties, cost, and performance under irradiation. Results of published work were reviewed on the effects of irradiation on beryllium oxide and beryllium oxide - uranium dioxide dispersions. A research proposal for a Maritime Gas-cooled Reaotor Moderator is included. (J.R.D.)
Date: September 11, 1959
Creator: Wallace, W. P. & Simnad, M. T.
Partner: UNT Libraries Government Documents Department

Effect of Massive Neutron Exposure on the Distortion of Reactor Graphite

Description: Distortion of reactor-grade graphites was studied at varying neutron exposures ranging up to 14 x 10/sup 21/ neutrons per cm/sup 2/ (nvt)/sup */ at temperatures of irradiation ranging from 425 to 800 deg C. This exposure level corresponds to approximately 100,000 megawatt days per adjacent ton of fuel (Mwd/ At) in a graphite-moderated reactor. A conventionalcoke graphite, CSF, and two needle-coke graphites, NC-7 and NC-8, were studied. At all temperatures of irradiation the contraction rate of the samples cut parallel to the extrusion axis increased with increasing neutron exposure. For parallel samples the needle- coke graphites and the CSF graphite contracted approximately the same amount. In the transverse direction the rate of cortraction at the higher irradiation temperntures appeared to be decreasing. Volume contractions derived from the linear contractions are discussed. (auth)
Date: May 28, 1963
Creator: Helm, J. W. & Davidson, J. M.
Partner: UNT Libraries Government Documents Department