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Experimental Stress Analysis of EGCR Pressure Vessel

Description: Report regarding the structural evaluations made of the upper head of the Experimental Gas Cooled Reactor (EGCR). This report includes descriptions of the model, experimental procedures, and analyses of the results. Appendix begins on page 139.
Date: 1961
Creator: Greenstreet, B. L.; Holland, R. W.; Maxwell, R. L.; Witt, F. J.; Shobe, L. R. & LaVerne, M. E.
Partner: UNT Libraries Government Documents Department
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Experiments and Analysis for SM-1 Core II With Special Components

Description: Abstract: This technical report contains a summary of analytical and experimental work performed on SM-1 Core II, with special components is presented. The effects of these special assemblies upon power distribution and core reactivity were calculated and compared to experimental measurements. A thermal analysis was conducted to determine steady state and transient performance of the special test components of the core as well as some of the hotter standard Core II components. Experimental work discussed includes individual reactivity effects of all the special elements and the total effect of all of the elements. Power mappings were also made and are reported.
Date: January 1, 1961
Creator: Lee, D. H.; Robinson, R. A. & Segalman, I.
Partner: UNT Libraries Government Documents Department
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Steam-Cooled Power Reactor Evaluation: Study of 300 Mw(e) Once-Through Superheater Reactor

Description: From introduction: "This report presents a conceptual design study of the once-through superheat reactor design based on experimental heat transfer results obtained during the fall of 1960 on Task F-2 of the AEC sponsored Nuclear Superheat Project."
Date: January 1961
Creator: U.S. Atomic Energy Commission
Partner: UNT Libraries Government Documents Department
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Superposition of Forced and Diffusive Flow in a Large-Pore Graphite

Description: Report describing an "investigation of steady-state counter-flow of gases in a large pore graphite" by exposing it to streams of helium and argon.
Date: 1961
Creator: Evans, R. B., III; Truitt, J. & Watson, G. M.
Partner: UNT Libraries Government Documents Department
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Thermonuclear Project Semiannual Progress Report: for Period Ending July 31, 1960

Description: From Introduction: "The present semiannual progress report introduces more active consideration than has been given in the past (by us, at least) to the importance of energy degradation in DCX-like devices. The following pages introduce also a new major apparatus designated as DCX-2. Some of these concepts cannot as yet be released publicly and therefore Chap. 3 of this report, covering the DCX-EP-B design, is being issued separately, with restricted distribution, as ORNL-3044."
Date: January 13, 1961
Partner: UNT Libraries Government Documents Department
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Evaluation of Kanigen, Electroless Nickel Plating for Steam Side of a Sodium Component Steam Generator

Description: Introduction: This is a final report on the evaluation of Kanigen electroless nickel plating for surfaces in contact with water and steam i a sodium heated AISI Type 316 stainless steel steam generator. The purpose of the coasting was to afford protection from stress corrosion cracking originating on the water-steam side of the unit. It has been concluded that the kanigen coating does not afford adequate protection for the services condition intended. This work was performed as part of the research and development program for the United States Atomic Energy Commission sodium Components Design Project.
Date: February 15, 1961
Creator: Alco Products (Firm).
Partner: UNT Libraries Government Documents Department
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Fission Product Activity in SM-1 Core I Primary System and Surface Contamination on SM-1 Type Fuel Elements. Task XVIII, Phases 2 and 3

Description: Abstract; The fission product data obtained during SM-1 Core I operation (June 1957 - May 1960) is reviewed briefly and interpreted. Evidence is presented to indicate that a fuel element defect was responsible for the high fission product activity level observed in the primary coolant. Relative escape coefficients are calculated and the defect size estimated. Anticipated fission product levels during SM-1 Core II and SM-1A Core I operation are estimated from alpha surface contamination data on completed fuel elements. The importance of in-line sampling for monitoring fission product activity is stressed as well as the need for failed fuel element detection methods.
Date: February 28, 1961
Creator: Hasse, Robert A. & Zegger, John L.
Partner: UNT Libraries Government Documents Department
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Final Design Report: DR-1 Gas Loop

Description: Report describing the performance, fission product tolerance, design, and costs of the DR-1 Gas Loop, which is an in-reactor test facility.
Date: March 1961
Creator: Baars, R. E.
Partner: UNT Libraries Government Documents Department
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Suitability of Inconel for Corrosion Protection on Water Side of Sodium Component Steam Generator

Description: Abstract; The heat exchanger and steam generator for the U.S. Atomic Energy Commission Sodium Components Project will be constructed entirely of type 316 stainless steel. Because of the susceptibility of this alloy to stress corrosion cracking, it is proposed to clad all areas of the steam generator with Inconel where the stainless steel will be exposed to water and steam. This report includes a discussion of the work by numerous investigators that justify the selection of Inconel for this service. A discussion of Inconel type welding alloys is also included.
Date: March 1, 1961
Creator: Phillips, Laurence E. & Vawter, Frank J.
Partner: UNT Libraries Government Documents Department
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The Cold Pressing of Sinterable UO₂

Description: The intent of this work was to explore more fully the pressing of sinterable UO2 powders into cylindrical compacts in the hope that a more precise prediction of green density in terms of powder properties, pressure, and geometry could be evolved.
Date: March 27, 1961
Creator: Levey, R. P., Jr.
Partner: UNT Libraries Government Documents Department
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Hazards Report for the SM-1 Core II With Special Components

Description: Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II and special components. The special components consist of impact specimens, a boron gradient rod, SM-2 elements, a PM-1-M element, and high burnup SM-1 Core I elements. The change in hazards, due to operation of SM-1 with Core II and the special components is evaluated. The analysis indicates there is no change in hazards.
Date: March 30, 1961
Creator: Coombe, J.; Lee, D.; Segalman, I. & Robertson, R.
Partner: UNT Libraries Government Documents Department
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Hazards Summary Report for the SM-1 Core Temperature and Flow Instrumentation: Task XIV

Description: Abstract; This technical report describes the changes in the SM-1 incurred by the experiment, Core Temperature and Flow Instrumentation (Task XIV), and evaluates the possible hazard involved in these changes. Temperature and flow measurements will be taken on a Task XIV instrumented stationary fuel element, instrumented control rod fuel element and other selected points in the SM-1 core to provide data on the core steady state and transient performance. The hazards evaluation consists of a nuclear evaluation, thermal and hydraulic analysis, description of tests to be performed, and discussion of containment integrity and maximum accident considerations.
Date: March 30, 1961
Creator: Coombe, J. R.; Brondel, J. O.; Lee, D. H. & Matthews, F. T.
Partner: UNT Libraries Government Documents Department
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Runaway Analysis for a Gas Cooled Reactor

Description: From abstract: "This report presents an IBM digital computer program that solves numerically and simultaneously the equations that describe gross reactor behavior under runaway conditions."
Date: April 1961
Creator: Becker, Richard A.
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for the SM-1 Core II Without Special Components

Description: Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II without special components. The SM-1 Core II components were made to specifications very nearly identical to those of SM-1 Core I. The differences consist of europium absorber sections, internal europium flux suppressors in the control rod fuel elements, and low impurity cladding. Each of the SM-1 Core II components with the exception of the five absorber sections new in SM-1 Core I were subjected to a Zero Power Experiment at the Alco Critical Facility. The results of this experiment indicate that the SM-1 Core II will have nuclear characteristics very similar to that of the SM-1 Core I. Since SM-1 Core II will be operated with the same mode of rod control, in the same core support structure, and with the same primary coolant flow conditions, the thermal characteristics should be essentially identical to that of SM-1 Core I. Also, all kinetic characteristics of SM-1 Core II should be identical to those of SM-1 Core I. This report demonstrates that there is no increase in potential for a hazardous situation at SM-1 due to the replacement of SM-1 Core I by SM-1 Core II.
Date: April 19, 1961
Creator: Gallagher, J. G.
Partner: UNT Libraries Government Documents Department
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