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Analysis of Stresses in Bellows

Description: 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.
Partner: UNT Libraries Government Documents Department
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Valve Stem Freeze Seal for High-Temperature Sodium

Description: 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.
Partner: UNT Libraries Government Documents Department
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Operation and Analysis of a 3000 KW Liquid Metal Model Steam Generator

Description: 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.
Partner: UNT Libraries Government Documents Department
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Piqua Prototype Handling System

Description: 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.
Partner: UNT Libraries Government Documents Department
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An Advanced Sodium-Graphite Reactor Nuclear Power Plant

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

A Multichannel Digital Recording System

Description: 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.
Partner: UNT Libraries Government Documents Department
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Hazards Analysis of the Organic Moderated Reactor Experiment

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Thermal Cycling Tests of Porous Uranium

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

OMRE Fuel Removal and Shipping Equipment

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

OMR (Piqua) Unitized Control-Safety Rod Prototype Tests

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Final Safeguards Summary Report for the Piqua Nuclear Power Facility

Description: 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
Partner: UNT Libraries Government Documents Department
open access

Performance of HNPF Prototype Free-Surface Sodium Pump

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

SRE Fuel Element Damage: An Interim Report

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Design Modifications to the SRE during FY 1960

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Thermal Cycling and Leakage Tests of 12-inch Valves for Sodium Service

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Metallurgical Aspects of SRE Fuel Element Damage Episode

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

300,000-KWE SGR Nuclear Power Plant of Current Technology

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Investigations of Neutron Penetration in TiH and Steel Slabs

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Corrosion and Activity Transfer in the SRE Primary Sodium System

Description: 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.
Partner: UNT Libraries Government Documents Department
open access

Sodium Reactor Experiment Power Expansion Program: Heat Transfer Systems Modifications

Description: 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.
Partner: UNT Libraries Government Documents Department
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