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Effect of Heat Flux on the Corrosion of Aluminum by Water. Part Iii. Final Report on Tests Relative to the High-Flux Isotope Reactor

Description: The effect of very high heat fluxes on the corrosion of 1100 and 6061 aluminum alloys by water was investigated. The test conditions generally simulated those expected to exist during operation of the High-Flux lsotope Reactor. At heat fluxes between 1 and 2 x l0/sup 6/ Btu/hr-ft/sup 2/ and with coolant temperatures and velocities in the ranges of 13l to 250 deg F and 3l to 51 fps, respectively, a layer of boehmite ( alpha Al/sub 2/O/sub 3/- H/sub 2/0), which has low thermal conductivity, forme… more
Date: December 20, 1961
Creator: Griess, J. C.; Savage, H. C.; Rainwater, J. G.; Mauney, T. H. & English, J. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Reactor Development Program Progress Report, August 1961

Description: Progress is reviewed on the following reactors: EBWR; Borax-V; ZPR-III- ZPR-VI; ZPR-IX; EBR-I; and EBR-II. An outline of fast and slow reactor safety studies in TREAT is presented. Progress is also reported in applied nuclear and reactor physics; development of reactor fuels, materials, and components; heat engineering technology; separation processes; and advanced reactor concepts. (T.F.H.)
Date: September 15, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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EXPERIMENTAL MEASUREMENTS OF THE SUCTION HEAD REQUIRED BY THE HALLAM PROTOTYPE FREE SURFACE SODIUM PUMP

Description: Hydraulic tests were made on the Hallam Prototype Free-Surface Sodium Pump to determine the net positive suction head (NPSH) required at various sodium flow rates. Pump performance data were also collected. The results indicate that an NPSH of 22 ft sodium is required at the design flow rate of 7200 gpm at approximates 1000 deg F, agreeing with computed values, and that the pump is designed with a safety margin of slightly over l0%. (D.L.C.)
Date: July 25, 1961
Creator: Atz, R.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Theoretical Study of the Transient Operation and Stability of Two-Phase Natural Circulation Loops

Description: Mathematical models of the time-dependent behavior of two-phase natural- circulation loops were used to predict the operation and to explain the unusual instability sometimes observed. The initial results obtained for a loop similar to the Univ. of Minnesota loop were used to formulate a more complex and accurate model, and the predicted transient behavior was in close agreement with the experimental results from the Minnesota loop. For a 300psia, high-pressure loop, unstable oscillatory behavi… more
Date: June 1, 1961
Creator: Garlid, K.; Amundson, N. R. & Isbin, H. S.
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries Government Documents Department
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Steady State and Transient Thermal and Hydraulic Analysis of SM-2 Termination Report

Description: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during-a loss of flow acident, ART-02, a onedimensional code, was used. This analysis indicated the SM-2 core is safe from burnout under steady state operation at design power level (28 Mw(t)) because no nucleate boiling exists, and the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of … more
Date: September 1, 1961
Creator: Segalman, I. & Bradley, P. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Heat Transfer With Laminar Flow in Concentric Annuli With Constant and Arbitrary Variable Axial Wall Temperature

Description: An analysis has been performed to determine the heat transfer characteristics for laminar forced-convection flow in a concentric annulus with prescribed surface temperatures. Three distinct problems were considered: (a) wall temperature prescribed at both the inside and outside wall; (b) inside wall temperature prescribed and the outside wall insulated; and (c) inside wall insulated and outside wall temperature prescribed. The solution for temperature distribution was similar to that obtained b… more
Date: December 1, 1961
Creator: Viskanta, R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Process for the Recovery of Uranium From Nuclear Fuel Elements Using Fluid-Bed Drying and Volatility Techniques

Description: A process scheme for the recovery of uranium from fuel elements has been developed. The scheme combines continuous fluid-bed drying and fluoride volatility techniques after initial dissolution of the fuel element in the appropriate aqueous system, hence the designation ADF, Aqueous Dry Fluorination Process. The application of this process to the recovery of uranium from highly enriched, low uranium-zirconium alloy plate-type fuels is described. ln the process, the feed solution is sprayed horiz… more
Date: November 1, 1961
Creator: Levitz, N.; Barghusen, J.; Carls, E. & Jonke, A. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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GAS-COOLED REACTOR PROGRAM. QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JUNE 30, 1961

Description: Activities are discussed for research in design investigations, and materials development and testing conducted in connection with the development of the EGCR. The discussions are given in terms of: reactor physics; reactor design studies; heat transfer and fluid now investigations; materials development; in- pile and out-of-pile testing of components and materials; and development of test loops and components. (B.O.G.)
Date: August 28, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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POWER-TO-VOID TRANSFER FUNCTIONS

Description: Variations in the distribution of steam bubble, the "void" distribution, in a boiling channel as a function of changes in heating power were studied. A rectangular test tube, of 1.11 x 4.44-cm cross section and 127-cm height, was inserted in a forced-circulation pressure loop. The tube was heated by passing an a-c current through the tube walls. A power oscillator was built which could give a 10% peak-topeak sinusoidal power modulation at any frequency in the interval from 0.01 to 10 cps. Varia… more
Date: July 1, 1961
Creator: Christensen, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Single Element Flow Tests for Type 3 (SM-2) Fuel Elements in SM-1, SM-1A, and PM-2A Cores

Description: Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized … more
Date: November 27, 1961
Creator: Krause, P. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Quarterly Status Report on LAMPRE Program for Period Ending August 20, 1961

Description: BS>All basic experiments planned for the LAMPRE I startup program were completed. The tests included operation at 100 and 200 kw utilizing both normal and half coolant flow, and full flow operation at 400 kw. At each power level, transfer function measurements were made, and a continuous run of approximately 60 hr duration was carried out to determine the characteristics of the reactivity loss first observed at 50 kw. Various power demand tests were made. Fuel and container development was cont… more
Date: September 1, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of the Initial Nuclear Superheat Critical Experiments. Supplementary Study Related to Bonus and Nuclear Superheat Programs

Description: A critical experiment program is carried out in a configuration similar to the BONUS reactor. The results give information concerning: the effects of different boilersuperheater geometries; the reactivity changes associated with superheater voiding or flooding; power regulation between the boiler and superheater regions; epithermal transmission probabilities for B-stainless steel and Cd control rods; the power flattening characteristics; and void simulation properties. The calculational methods… more
Date: January 30, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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COOLING OF THE HFIR BERYLLIUM REFLECTOR FOLLOWING A REACTOR SCRAM OR AN ELECTRICAL POWER OUTAGE

Description: Thermal stresses in the HFIR beryllium reflector were computed for the unlikely case where the reactor is scrammed with a simultaneous loss of coolant flow and for the case following an electrical power outage where the reactor power level and the coolant flow rate are reduced simultaneously. For the case where the reactor is scrammed with a sudden loss of the coolant flow, the resulting maximum tensile thermal stress following the scram is 22,500 psi. In case of an electrical power outage, the… more
Date: December 12, 1961
Creator: McLain, H. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Radiological Physics Division Semiannual Report, January-June 1961

Description: Twenty papers are presented on various projects pursued in the Radiologmcal Physics Division. The topmcs of the papers include radioactivities of bone and air, cesium137 content in human subjects, liquid scintillators, fluid flow, and air and soil temperature cycles. Separate abstracts were prepared for 16 of the papers. (D.L.C.)
Date: September 1, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Supporting Analysis and Derivation of Dimensional Tolerance Specifications for Core II of SM-1A & PM-2A

Description: A method is presented for translating inspection measurements of fuel plate spacing to obtain minimum coolant channel clearances under reactor operating conditions. Considerations of fuel plate ripple growth and the inspection procedure used are included. The method is applied to establish dimensional tolerance specifications used for the procurement of SM-1A and PM-2A Core II. (auth)
Date: November 1, 1961
Creator: Brondel, J. O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

CORE REMOVAL COOLING SYSTEM-SECTION II. CORE I, SEED I. Test Results T- 641113. Section 2

Description: A test was performed on June 19, 1959 to determine the capacity of the Core Removal Cooling System for removing reactor decay heat under split-flow'' conditions. The system operated satisfactorily during this test; the pumps developed a flow of approximates 73 gpm at a total head of 254 ft water, as compared with their rated capacity of 75 gpm at a total head of 250 ft water. (D.L.C.)
Date: May 19, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fuel Cycle Program, a Boiling Water Reactor Research and Development Program. Third Quarterly Report, January 1961-March 1961

Description: The continuing analysis of the VBWR core resulted in refinements in the calculations for reactivity in voids, flux leakage and resonance escape probability. The Zircaloy cladding for 25 fuel assemblies was received and passed inspection. Preliminary measurements of VBWR flux oscillations, used to develop instrumentation and data interpretation techniques, showed random normally-distributed oscillations with a predominant frequency of 0.5 to 1.0 cycles/second. A model for analog computer simulat… more
Date: October 31, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Pressure Drop of Multirod Elements With Helical Spring Spacers

Description: The pressure drop of a new fuel element design concept of spacing rods by means of helical wire springs was investigated experimentally and analytically. Extensive single- and two-phase pressure drop data at 1,000 psia were obtained for one flow geometry and helical spring spacer. Test conditions ranged from 0.7 to 1.2 x 10/sup 6/ lb/hr ft/sup 2/ in mass velocity and from 0 to 15% in quality. The effect of the specific spring which was tested was to increase the over-all pressure drop by 70%. A… more
Date: June 1, 1961
Creator: Quinn, E. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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DEVELOPMENT OF CENTRIFUGAL COMPRESSORS

Description: Coolant flow for gas-cooled in-pile loops must be supplied during irradiation test runs. A centrifugal compressor has been designed and developed for circulating helium at volume flows from 75 to 250 acfm at compressor suction conditions of 400 psi and 600 deg F. The compressor using grease-lubricated ball bearings has operated satnksfactorily for a total of 3500 hr. (auth)
Date: October 1, 1961
Creator: Namba, I.K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Feasibility Study of a New Mass Flow System. Quarterly Report No. 5, June 1, 1961 to August 31, 1961

Description: Activities are reported on development work on a mass flow system capable of measuring externally the properties of homogeneous flow, slurries, highly corrosive fluids, and multi-phase fiuids. In the proposed system, the fluid passes through an S-shaped tube wherein measurements of angular momentum and density yield mass flow directly. (B.O.G.)
Date: September 20, 1961
Creator: Haffner, J. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SUMMARY OF HRT RUN 21

Description: The HRT was operated experimentally during run 2l at powers up to 5 Mw to explore the limiting conditions of fuel stability and to demonstrate the reliability of the system. The effect of core pressure on fuel stability was investigated over the range from l250 to 1750 psig. Stable operation at 5 Mw (2.6 Mw in the core) was demonstrated at 1250 psig. At 1600 and 1750 psig, fuel instability accompanied by rapid loss of reactivity occurred at powers down to 2.5 Mw. The threshold power for reactiv… more
Date: October 10, 1961
Creator: Haubenreich, P.N.; Bauman, H.F.; Bradley, N.C.; Engel, J.R.; Kolb, J.O.; Piper, H.B. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Engineering Capabilities of in-Pile Irradiation Facilities Used by GE-ANPD

Description: GE-ANPD had exclusive use of irradiation testing facilities in four reactors: the Materials Test Reactor (MTR), the Engineering Test Reactor (ETR), the Low lntensity Test Reactor (LlTR), and the Oak Ridge Research Reactor (ORR). A compilation of data concerning the GE-ANPD facilities in these reactors is presented. lnstrumentation capabilities, dimensions of the in-pile tubes, flow characteristics, control capabilities, and detailed design data are listed. (auth)
Date: October 1, 1961
Creator: Harry, R. J. & Fries, R. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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