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Part I. Theory report for CREEP-PLAST computer program: analysis of two- dimensional problems under simultaneous creep and plasticity

Description: >Solution of the combined creep and plasticity problem is considered. Both creep theories, the equation-of-state and memory theories, are used within the framework of the finite-element computatioral method. Because of the limited creep data, which is available only for single-step simple extension experiments, the integral expansion in the memory theory formulation is limited to one nonlinear superposition integral. However, instead of using direct superposition as proposed earlier, the kernel function was derived as an integral transformation that reduced the time--temperature- stress relationship to a single quantity. The equationof-state formulation is based on the monotoric strain-hardening rule app1ied to the primary creep componert only. The uniaxial stress- strain law is first discussed with emphasis on the memory theory. Then the incremental stress- strain relations for the combined creep and plasticity problem are derived for both creep theories. The time--temperature- stress transformation is derived for a particular class of material that exhibits well- defined primary and secondary creep parts such as stainless steel. The properties of the resulting stress- strain integral are discussed in relation to the direct-superposition method. Finally, example analyses are given. (auth)
Date: January 1972
Creator: Rashid, Y. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Liquid Metal Fuel Reactor Experiment (Phase 1C): Supplement to LMFRE Reference Design Report

Description: This supplement describes a number of modifications to the Liquid Metal Fuel Reactor Experiment (LMFRE), including (1) engineering design changes to achieve economies in the original reference design, and (2) a revised cost estimate that reapportions the various construction, research and development, and operating costs in accordance with procedures established by the Atomic Energy Commission.
Date: December 1957
Creator: Babcock & Wilcox Company
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Oak Ridge Metals and Ceramics Division Annual Progress Report: 1963

Description: Report documenting research and developments made by the Metals and Ceramic Division of the Oak Ridge National Laboratory. Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other program activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)
Date: November 11, 1963
Creator: Oak Ridge National Laboratory. Metals and Ceramics Division.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Reduction of Liquid Radioactive Wastes to Solids: Scouting Tests

Description: Abstract: Scouting experiments indicated that calcination of highly active Darex and Sulfex decladding and Purex extraction wastes will not release hazardous amounts of fission products to the atmosphere. Simulated wastes containing up to 90 curies/liter or activity were evaporated and calcined to 750°C and the off-cases passed through a condenser and a series of caustic scrubbers. Less than 0.1% of the fission products was released, of which ruthenium was 10-90%. Fission product release was lower from neutralized than from acidic wastes. Fission products were leached appreciably, 0.05% of the beta and 0.02% of the gamma, from the calcination solid with water in 96 hr.
Date: November 24, 1958
Creator: Hancher, C. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Gas-Cooled Reactor Project Quarterly Progress Report: June 1960

Description: Report documenting ongoing research and developments at the Oak Ridge National Laboratory's Gas-Cooled Reactor Project. Design Investigations: The effects on the power distribuestablished. A mathematical model was developed for studying shifting of the coolant stream as it moves along a rod in order to predict the temperatures of the parallel streams as they progress through the reactor. A fuelelement life code developed for computing the internal temperature structure, the amount of fission gas released, the internal pressure, the cladding strain when the internal pressure exceeds the coolant pressure, and the creep damage was used for comparing top-loading and inventedloading fuel programs for the EGCR. A statistical method was developed for estimating the probability that the hot spot on the EGCR fuel element will exceed a given temperature. A method of cooling the EGCR control rods was developed that will minimize diversion of coolant flow through leakage paths between graphite blocks. A preliminary design of a control rod cooled by this method was developed. Means for reducing the thermal stresses in the top head nozzles of the EGCR pressure vessel were studied. The stresses in the graphite sleeves of the EGCR fuel elements were calculated, and the maximum stress was found to be within the allowable limit. A study was made of the thermal stresses in the EGCR pressure-vessel support skirt, and a satisfactory design was developed. Procedures for removing ruthenium and cerium contamination from steel were outlined and incorporated in procedures for decontaminating the EGCR charge and service machines. Experimental information was obtained on the thermal characteristics of the specified EGCR fuel cluster. The effect of relative orientation of adjacent clusters on the heattransfer distribution in the downstream element was studied by means of mass-removal measurements on naphthalene-coated reds. Velocity distributions in the downstream element of two adjacent EGCR-type clusters ...
Date: August 22, 1960
Creator: Oak Ridge National Laboratory
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Heat Transfer Correlation for Finned Casks

Description: Design of finned casks for dissipation of heat from radioactive decay usually requires reliance on generalized correlations in the literature which do not necessarily apply to the specific cask design. A correlation was developed, based on temperature profile measurements, for the design of upright cylindrical casks with vertical fins for convective and radiant heat transfer to ambient air. Temperature data at various heat loads were obtained for two different cask sizes of the same basic design. Each cask is mounted on a steel pallet and contained within a steel mesh cage. The smaller cask, which has 23 fins, has been approved (DOT-SP-6321) for shipment of up to 1400 W (th), and approval is being obtained (AEC AL USA/9503 BLF) for shipment of up to 3500 W heat load in the larger, 60-fin cask. The applicable theoretical equations were fit to the temperature data for both casks by simply adjusting the value used for the number of fins. The resulting correlation provides a reliable method for interpolation and extrapolation and for design of similar finned casks.
Date: April 1, 1974
Creator: Griffin, J. F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Molten-Salt Reactor Program Semiannual Progress Report for Period Ending February 28, 1965

Description: Report issued by the Oak Ridge National Laboratory discussing semiannual progress made by the Molten-Salt Reactor Program. Progress of operations, construction, engineering analysis, and development is presented. This report includes tables, illustrations, and photographs.
Date: June 1965
Creator: Briggs, R. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Scale Control With Graphite Heat-Transfer Tubes of Controlled Permeability to Steam

Description: From Introduction: "The present project authorized under OSW Contract No. 14-01-0001-685 was undertaken to investigate the feasibility of controlling calcium sulfate scale by permission of condensate through graphite heat-transfer tubes. Specifically, the following tasks were accomplished during the course of this contract: Graphite tubes 108" long X 2" O.D. x 1 1/2" I.D. were manufactured at three permeability levels. A test apparatus was constructed and proved out as a useful and reliable probe. Tests were conducted at two brine concentrations. Some correlation of test considerations with scale formation was found. The economic practicality and the feasibility of using permeable tubes for desalinization were established."
Date: July 1967
Creator: Vaaler, L. E. & Hulswitt, C. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department