Resource Type
Serial/Series Title
Decade
Month
168 Matching Results
Results open in a new window/tab.
Analysis of Stresses in Bellows
Abstract: Design charts and systematic design forms are presented for simplified calculations to check the number of convolutions and thickness required to limit the deflection and pressure stress range in three types of bellows.
Date:
October 15, 1964
Creator:
Anderson, W. F.
System:
The UNT Digital Library
Engineering Evaluation of a Mixed Alloy Fuel Element Irradiated at Elevated Temperatures in the SRE
Abstract: A fuel material evaluation was made by destructively examining a full-scale experimental fuel element, irradiated in the SRE to a maximum of 850 Mwd/MTU.
Date:
June 1, 1960
Creator:
Ballif, J. L.; Hayward, B. R. & Walter, J. W.
System:
The UNT Digital Library
Valve Stem Freeze Seal for High-Temperature Sodium
Abstract: An experimental study of valve stem freeze seals was undertaken as part of the effort to develop components for high temperature service in advanced sodium-cooled reactor systems.
Date:
July 30, 1960
Creator:
McDonald, J. S.
System:
The UNT Digital Library
Operation and Analysis of a 3000 KW Liquid Metal Model Steam Generator
Abstract: A 3000 kw (thermal) bayonet duplex tube model steam generator was performance-tested in a liquid metal test loop at MSA Research Corporation, Callery, Pennsylvania, under the cognizance of Atomics International.
Date:
February 28, 1961
Creator:
Webster, L. J.
System:
The UNT Digital Library
Piqua Prototype Handling System
Abstract: Equipment has been developed to handle the fuel elements and control rods for the Piqua (OMR) reactor.With the handling machine, which consists of a shielded cask mounted on a gantry, a fuel element can be replaced in the core in about 27 minutes.
Date:
May 1960
Creator:
Nadler, H.
System:
The UNT Digital Library
An Advanced Sodium-Graphite Reactor Nuclear Power Plant
Abstract: This report describes an advanced sodium-cooled, graphite-moderated nuclear power plant which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency.
Date:
March 15, 1960
Creator:
Churchill, J. R. & Renard, J.
System:
The UNT Digital Library
A Multichannel Digital Recording System
Abstract: This report is a description of a 200 channel digital recording system used to record high temperature strain gage outputs and associated temperatures.
Date:
September 15, 1960
Creator:
Truitt, R. W.
System:
The UNT Digital Library
Hazards Analysis of the Organic Moderated Reactor Experiment
Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.
Date:
December 15, 1959
Creator:
Williams, R. O., Jr.; Allen, W. O.; Ash, E. B.; Scott, W. W.; Shimazaki, T. T.; Sletten, H. L. et al.
System:
The UNT Digital Library
Thermal Cycling Tests of Porous Uranium
Abstract: It has been proposed that a fuel element containing porous uranium be used for moderately high temperature applications. In an effort to determine the structural integrity of this type of fuel body, thermal cycling tests have been conducted on porous uranium.
Date:
March 15, 1960
Creator:
Wilkinson, L. E.
System:
The UNT Digital Library
OMRE Fuel Removal and Shipping Equipment
Abstract: A portable assembly for handling and shipping OMRE fuel elements is described and details of its operation are given. Problems of heat transfer and radiation shielding are discussed, and detailed analysis are presented.
Date:
April 1, 1960
Creator:
Mallon, P. J.; Duncan, D. S. & Noyes, R. C.
System:
The UNT Digital Library
Fabrication Modification Development for OMRE Third Core Loading
Abstract: This report describes the fabrication of elements for the OMRE third core loading.
Date:
July 15, 1961
Creator:
Peters, E. & Binstock, M. H.
System:
The UNT Digital Library
OMR (Piqua) Unitized Control-Safety Rod Prototype Tests
Abstract: A unitized magnetic jack driven control-safety rod has been developed for the 45.5 thermal megawatt organic moderated reactor (Piqua).
Date:
June 30, 1960
Creator:
Howell, J. D. & Weeks, C. C.
System:
The UNT Digital Library
11,400 KW Nuclear Power Plant Employing an Organic Moderated Reactor: Preliminary Description
Abstract: The preliminary design is described for a small electric-power-generating plant powered by an organic moderated reactor. System and component requirements are discussed and possible design configurations and equipment are described.
Date:
1957
Creator:
Wheelock, C. W.
System:
The UNT Digital Library
Final Safeguards Summary Report for the Piqua Nuclear Power Facility
Summary: This report contains a description of the final design of the Piqua Nuclear Power Facility (PNPF); an outline of the test and operating procedures, and the organization and responsibilities; and a summary of the hazards and safeguards analyses that have been conducted to evaluate the safety of the facility operations.
Date:
August 1, 1961
Creator:
unknown
System:
The UNT Digital Library
Performance of HNPF Prototype Free-Surface Sodium Pump
Abstract: A free-surface centrifugal pump, incorporating a hydraulic bearing running in sodium, was operated at the conditions required for service in the HNPF.
Date:
1960
Creator:
Atz, R. W.
System:
The UNT Digital Library
SRE Fuel Element Damage: An Interim Report
Abstract: During the course of power run 14 on the Sodium Reactor Experiment (SRE) at low power, the temperature difference among various fuel channels was found to be undesirably high. Normal operating practices did not succeed in reducing this temperature difference to acceptable values and on July 26, 1959, the run was terminated.
Date:
November 30, 1959
Creator:
Jarrett, A. A.
System:
The UNT Digital Library
Design Modifications to the SRE during FY 1960
Abstract: The means to prevent the recurrence of tetralin leakage into the SRE sodium systems are discussed. Included is a description of the redesign of system components to utilize alternate coolants such as nitrogen, air, and NaK.
Date:
February 15, 1961
Creator:
Deegan, G. E.; Dermer, M. D.; Flanagan, J. S.; Gower, G. C.; Hall, R. J.; Hinze, R. B. et al.
System:
The UNT Digital Library
Thermal Cycling and Leakage Tests of 12-inch Valves for Sodium Service
Abstract: Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of commercially available valves considered for use in the sodium coolant system of the Hallam Nuclear Power Facility.
Date:
May 1, 1960
Creator:
Baroczy, C. J.
System:
The UNT Digital Library
Metallurgical Aspects of SRE Fuel Element Damage Episode
Abstract: An investigation of the metallurgical aspects of the SRE fuel element episode, that occurred July 26, 1959, has been completed.
Date:
October 15, 1961
Creator:
Ballif, J. L.
System:
The UNT Digital Library
300,000-KWE SGR Nuclear Power Plant of Current Technology
Abstract: This report describes a 300,000-kwe, sodium-cooled, graphite-moderated nuclear power plant based on existing technical information.
Date:
August 1, 1960
Creator:
Renard, J.; Peckinpaugh, C. L. & Aronstein, R. E.
System:
The UNT Digital Library
Investigations of Neutron Penetration in TiH and Steel Slabs
Abstract: A multigroup P1 approximation for hydrogen scattered neutrons has been developed and applied to the study of neutron flux distributions in titanium hydride and steel shield systems.
Date:
May 15, 1961
Creator:
Karcher, R. H.
System:
The UNT Digital Library
Corrosion and Activity Transfer in the SRE Primary Sodium System
Abstract: An evaluation extending over a two-year period was made of primary system sodium and of stainless steel, zirconium, and beryllium specimens exposed in the hot and cold legs of a bypass loop in the primary system of the Sodium Reactor Experiment (SRE).
Date:
October 30, 1961
Creator:
Johnson, H. E.
System:
The UNT Digital Library
Sodium Reactor Experiment Power Expansion Program: Heat Transfer Systems Modifications
Abstract: Under the Power Expansion Program (PEP), modifications have been made to the Sodium Reactor Experiment (SRE) facility to improve plant reliability and permit an increase in power to 30 Mwt, with a reactor coolant outlet temperature up to 1200°F.
Date:
October 9, 1964
Creator:
Freede, W. J. & Roberts, J. K.
System:
The UNT Digital Library
Operating Experience with Heat Transfer System Pumps at the Hallum Nuclear Power Facility
Introduction: It is the purpose of this report to describe the operating and maintenance experience obtained at HNPF on the sodium heat transfer pumps.
Date:
June 15, 1964
Creator:
Durand, R. E.
System:
The UNT Digital Library