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Experimental Studies of Transient Effects in Fast Reactor Fuels

Description: An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO₂ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F.… more
Date: November 15, 1962
Creator: Field, J. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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NS Savannah Upgrading Analog Computer Analysis

Description: From abstract: "This work amplified previous design studies by considering accidents of greater severity, automatic limiting devices to preclude safety action, and more realistic parameters based on operating experience."
Date: November 1962
Creator: Rosser, R. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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EOCR Control Rod and Driver Fuel Hydraulic Tests

Description: Report discussing the results of experiments on an EOCR prototype control rod, control rod drive, and driver fuel assembly that were extensively tested in 500 deg F Santowax and 132 deg F water. These tests not only established the operating characteristics of these assemblies but also revealed a number of deficiencies in the various components. It includes descriptions of the control rod, control rod drive, and driver fuel assembly together with descriptions of their method of operation.
Date: October 19, 1962
Creator: Harrison, L. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effects of Major Parameters on Cycle Efficiency and Cost for a Gas-Cooled Reactor Turbine Power Plant

Description: Cycle study performed on a closed loop Brayton cycle incorporating a gas-cooled nuclear reactor, regeneration, and compressor intercooling. The purpose of the study was to determine the effect of various cycle parameters on efficiency and cost.
Date: October 5, 1962
Creator: Miskell, R. V.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Test Program for 30 Megawatt Prototype Sodium Intermediate Heat Exchanger and Steam Generator

Description: The designer's concept of a test program for the 30-Mw prototype intermediate heat exchanger and steam generator designed and fabricated as part of the Sodium Components Development Program is presented. The performance data will serve to verify the thermal design, or allow application of improved techniques to future designs, give an improved basis for stress analysis in design of future units, and demonstrate the capability and limitations of the units in relation to the performance specifica… more
Date: September 14, 1962
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Power Distributions for Type 3 Replacement Cores for SM-1, SM-1A and PM-2A

Description: Abstract: The most adverse power distribution and power peaking factors have been developed for Type 3 replacement cores for SM-1, SM-1A, and PM-2A. Also included is the 37 element Type 3 core for SM-1, which is known as SM-1 Core III and is a prototype of a Type 3 core for PM-2A. The power distributions were compared with measured distributions obtained on type 3 fuel elements in SM-1, SM-1A and PM-2A core configurations. Using the measured data as a standard, correction factors were obtained … more
Date: July 25, 1962
Creator: Oggerino, J. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Thermal Analysis of SM-1 Core III

Description: Abstract This technical report covers the thermal analysis performed on the SM-1 Core III for both steady state and transient conditions is reported. SM-1 Core III will be used as a test for Type 3 elements in a PM-2A Core. The steady state analysis indicated minimum departure from nucleate boiling ratios (DNBR) for both design and scram conditions above the minimum criteria of 1.5. Local nucleate boiling was noted in the hot internal channels and lattice passage at scram power conditions. Loss… more
Date: June 29, 1962
Creator: Davidson, S. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for PM-2A Core II

Description: Abstract: This technical report describes the changes incurred in the PM-2A by the planned insertion of PM-2A Core II and the replacement of the startup and check sources. PM-2A Core II components were fabricated to specifications very nearly identical to those of PM-2A Core I. The essential difference in the cores is the boron loading which permits PM-2A Core II to meet a "one-stuck rod criteria" at beginning of life. This core has been subjected to a zero power experiment and loading procedur… more
Date: June 21, 1962
Creator: Coombe, John R. & Stephenson, L. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Prediction of the Critical Heat Flux in Forced Convection Flow

Description: From summary: "A superposition model is developed to predict the critical heat flux in forced convection flow. The model is applied to available experimental results in boiling water flows and good agreement is obtained between the model and test data over the multitude of geometries, flow rates, pressures, and fluid enthalpies tested to-date."
Date: June 20, 1962
Creator: Levy, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

SM-1 Shielding Analyses

Description: Abstract: This technical report analyzes gamma dose rate and neutron measurements in their relation to the SM-1 shield design and is a continuation of previous shielding measurements and analyses reported in APAE-35 and APAE-35 Supplement 2. The data reported herein are spent fuel element and rod drive pit gamma dose rates. An analysis of gamma dose rates off the core midplane is presented and compared with test data.
Date: June 20, 1962
Creator: Stephenson, L. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Silver - Cadmium - Indium Absorber Development

Description: Abstract: This technical report covers development of an AG-Cd-In alternate absorber section for Army Type SM reactors. It describes the absorber material composition and the geometric configuration. It gives the nuclear and thermal analyses supporting this configuration and a detailed description of the manufacturing practice employed in fabricating the final design component.
Date: June 13, 1962
Creator: Shaw, R. A. & Harris, R. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

The Development and Testing of UO2 Fuel Systems for Water Reactor Applications: Summary Report, July 1, 1961 - June 15, 1962

Description: From introduction: "The work described in this report represents the Joint United States - European Atomic Energy Community effort which is in keeping with the spirit of cooperation in contributing to the common good by the sharing of scientific and technical information and minimizing the duplication of effort by the limited pool of technical talent available in Western Europe and the United States."
Date: June 1962
Creator: Murtha, B. E. & Chernock, W. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Simplified Power Conversion: Unit Study

Description: From abstract: "This report presents the results of a feasibility study on a simplified power conversion unit primarily for use with nuclear, steam power plants for military applications."
Date: June 1962
Creator: Clark, P. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Sodium Mass Transfer - I: Test Loop Design

Description: From abstract: "This report presents the design, fabrication, assembly, operating procedures, and start-up data for six experimental test loops to examine the effect of steel exposed to sodium at temperatures as high as 1300 F."
Date: June 1962
Creator: Lockhart, R. W.; Billuris, G. & Lane, M. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Army PWR Support and Development Program Six Months Summary Report : October 1, 1961 - March 31, 1962

Description: Abstract: Progress is reported on research and development tasks under the Program Plan for Engineering Support and Development of Army Pressurized Water Reactor Power Plants, Contract AT(30-1)-2639, during the six months' period October 1, 1061 to March 31, 1962.
Date: May 25, 1962
Creator: Dixon, M. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Comparative Cost Study of Processing Stainless Steel-Jacketed UO2 Fuel: Mechanical Shear-Leach vs Sulfex-Core Dissolution

Description: Comparison of the economics of mechanical shear-leach and Sulfex decladding-core dissolution head end treatments for processing typical tubular bundles of stainless steel-jacketed UO2 nuclear fuels.
Date: April 23, 1962
Creator: Adams, J. B.; Benis, A. M. & Watson, C. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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