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Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 4. Steam Driven Coolant Pumps

Description: Fourth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).
Date: October 23, 1961
Creator: Combustion Engineering, inc. Nuclear Division.
Item Type: Refine your search to only Report
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.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Criteria for Evaluating Hazards Involved in Proposed Tests On and/or Modifications To the SM-1

Description: Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.
Date: October 18, 1961
Creator: Scoles, J. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design Criteria for Irradiated Vessels Task 6.0 Summary Report

Description: Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the … more
Date: September 29, 1961
Creator: McLaughlin, D. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design, Fabrication, and Irradiation of Superheat Fuel Element SH-4B in VBWR

Description: From abstract: "The design, fabrication, and irradiation results are described for a 0.028 inch thick 304 stainless clad fuel element (SH-4B) exposed in the Vallecitos Boiling Water Reactor loop under superheat conditions."
Date: September 1, 1961
Creator: Spalaris, C. N.; Boyle, R. F.; Evans, T. F. & Esch, E. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effect of Radiation Damage on SM-1, SM-1A and PM-2A Reactor Vessels

Description: Report describing the status of the SM-1, SM-1A, and PM-2A reactors, specifically regarding the effects "of irradiation on nil-ductility transition temperature and the associated problem of brittle fracture." (p. iii)
Date: October 14, 1961
Creator: McLaughlin, D. W.; Rowekamp, B. J.; Chittum, R. A.; Coombe, J. R.; Kelleman, R. W.; Bobe, P. E. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Evaluation of Iron- and Nickel-Base Alloys for Medium and High Temperature Reactor Applications, Part 1

Description: Report discussing the evaluation of the "behavior of austenitic and ferritic stainless steels, iron-chromium-aluminum base alloys and nickel-base superalloys, in medium and high temperature simulated reactor environments" (p. 2).
Date: February 1961
Creator: Pessl, H. J.
Item Type: Refine your search to only Report
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 r… more
Date: February 15, 1961
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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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.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: January 1, 1961
Creator: Lee, D. H.; Robinson, R. A. & Segalman, I.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Extended SM-2 Critical Experiments : CE-2

Description: Abstract: This technical report contains a description and results of a second series of critical experiments performed on the SM-2 core mock-up, as additional to the first series of experiments reported in APAE No. 54. The SM-2 core mock-up contains 36.4 kg U-235 and and estimated 67.9 gm B-10. The equivalent diameter and the active height are about 22 in.; the metal-to-water volume ration is 0.344. Data is presented on activation, reactivity, and stuck rod measurements. All measurements were … more
Date: June 30, 1961
Creator: McCool, W. J.; Robinson, R. A.; Weiss, S. H.; Raby, T. M.; Schrader, E. W. & Walthousen, L. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Fabrication of Fuel Rods by Tandem Rolling

Description: From introduction: "The purpose of this report is to present the details of the exploratory and developmental work on tandem rolling, and the subsequent fabrication of tandem rolled fuel rods for irradiation testing in the Vallecitos Boiling Water Reactor."
Date: July 1961
Creator: Lingafelter, J. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Failure Test of a Double Chambered NaK-Filled Irradiation Capsule

Description: Report describing tests "in which a double chambered irradiation capsule containing NaK (sodium-potassium alloy) in its inner chamber was deliberately perforated to allow NaK (the heat transfer medium) and water to react within the capsule chambers" (p. 2). The report includes descriptions of the materials used in the tests, test procedures, and results.
Date: September 1961
Creator: Kosut, B. S.; Leggett, R. D. & Marshall, R. K.
Item Type: Refine your search to only Report
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 o… more
Date: February 28, 1961
Creator: Hasse, Robert A. & Zegger, John L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Hazards Evaluation of the SM-1 Penetrated Gasket

Description: Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.
Date: September 8, 1961
Creator: Coombe, J. R.; Gebhardt, F. G. & James, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for Insertion of the PM-1-M-2 Element in the SM-1 Core II

Description: Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as def… more
Date: September 1, 1961
Creator: Coombe, J. R.; Scoles, J. F.; Brondel, J. O. & Lee, D. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for SM-1 Core II With the SM-1 Core II High Burnup Elements Replaced with SM-1 Core I Spare Elements

Description: Abstract: The removal of both SM-1 Core I high burnup elements from the SM-1 Core II and the insertion of two SM-1 Core I spare elements i their places are discussed. Nuclear and thermal characteristics of Core II with the change are presented and conclusion related to the change in hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative… more
Date: October 9, 1961
Creator: Coombe, J. R.; Lee, D. H. & Matthews, F. T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for SM-1 Core II Without the SM-1 Core I High Burnup Elements and With the PM-1-M-2 Element

Description: Abstract: The removal of both SM-1 Core I high burnup elements from SM-1 Core II and the insertion of the PM-1-M-2 element and the SM-1 Core I spare element in SM-1 Core II is discussed. Nuclear and thermal characteristics of Core II with these changes are presented and conclusions related to the changes in the hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself… more
Date: October 7, 1961
Creator: Coombe, J. R.; Lee, D. H. & Mathews, F. T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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