UNT Libraries Government Documents Department - 1,888 Matching Results

Search Results

open access

A 23-Group Neutron Thermalization Cross Section Library

Description: A set of 23-group neutron cross sections for use in the calculation of neutron thermalization and thermal neutron spectral effects in SNAP reactors is compiled. The sources and methods used to obtain the cross sections are described. (auth)
Date: July 15, 1963
Creator: Doctor, R. D. & Boling, M. A.
open access

40-MW(e) Prototype High-Temperature Gas-Cooled Reactor Research and Development Program. Quarterly Progress Report for the Period Ending June 30, 1962

Description: Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two …
Date: October 31, 1963
open access

100-N technical manual. Volume 2A: Systems descriptions

Description: This report contains engineering drawings for the control room, reactor monitoring systems, and reactor control systems for the N reactor. Each console in the control room is detailed. Other systems discussed are: stack air monitoring system, charging machine control systems, and heating and ventilation control systems. A N reactor plant glossary is included.
Date: December 31, 1963
open access

190-H drawdown test

Description: A discrepancy of about 1000 gpm has existed between the full-flow recorded 190, 105 and ROL flows. While past operating practices have not used the 190 or ROL flow rates for official purposes, the disquieting, though not theoretically unexplicable, differences require some quantitative resolution. On November 24, 1962, a drawdown test of the 190-H storage tanks was performed to establish the accuracy of the various flowmeters. The drawdown test of the 190 storage tanks was run at the beginning of a scheduled reactor shutdown. With the full reactor flow supplied by the electric process pumps feeding from the storage tanks, the 183-H supply to the storage tanks was valved off. Additionally, non-process water usually taken from the storage tanks was valved off. The storage tank water levels were taken, then recorded as a function of time.
Date: January 17, 1963
Creator: Cremer, B. R. & Bokish, K. P.
open access

A 200-Watt Conduction-Cooled Reactor Power Supply for Space Application

Description: The limited supply of relatively long-half-life isotopes having a reasonably high power density and the low conversion efficiencies obtainable with thermoelectric devices have so far limited the power output of isotope-fueled sources of electric power to several tens of watts. In addition, the high cost of the available isotopes results in a very large expense for isotope-fueled generators producing several hundred watts. It appears that a small, minimumweight, conduction-cooled reactor is an attractive alternate to the isotope-fueled power supplies in the 200-w size range. The proposed reactor is a small, high-density fast core of U/sup 233/ surrounded by a beryllium reflector. This approach, generally speaking, gives a reactor that is more compact and of lighter weight than can be obtained with a moderated system having a softer neutron spectrum. In the reactor design, the path of heat flow is from the core to the inner reflector and then to the thermoelements in close contact with the inner reflector. The reject heat flowing from the thermoelement cold junctions enters the outer pontion of the reflector, which acts as the heat sink and conducts the reject heat to the large, circular, tapered-fin radiator which is attached to the reflector. Survey physics calculations for various reactor systems fueled with U/sup 235/, U/sup 233/, and Pu/sup 239/ are reported. Some limits imposed on the system design by the thermoelectric generator are discussed, and the problem of radiator design for the space environment is treated in some detail. No attempt is made to present a detailed final design of the power supply; rather, the report is restricted to a general delineation of the limits imposed by various parameters and a resulting final conclusion as to the performance limits of small conduction-cooled reactors in this size range. (auth)
Date: March 1, 1963
Creator: MacFarlane, D. R.
open access

1000 Mwe Closed Cycle Water Reactor Study

Description: This report has two volumes, volume 1 contains the summary and detailed description of plant design, volume 2 contains a comprehensive nuclear evaluation of the reactor core.
Date: March 1, 1963
Creator: Westinghouse Electric Corp., Pittsburgh, Pa. Atomic Power Div.
open access

1000 MWE Closed Cycle Water Reactor Study Volume II

Description: This report includes the nuclear evaluation that has been conducted for the purpse of studying those problem areas which are expected to increase in severity as the core size is increased to produce 1000 MWE.
Date: March 1, 1963
open access

1559/RE: A CODE TO COMPUTE RESONANCE INTEGRALS IN MIXTURES

Description: The computer program 1559/RE is an experimental IBM-704 code in FORTRAN language for computing the resonance integrals of isotopes in mixtures in the presence of hydrogenic moderation. There may be up to four isotopes, each with no more than 75 resolved resonance levels. Doppler broadening and interference scattering are included No estimate is made of contributions from unresolved resonances. Typical running times are 30 min (with no Doppler broadening) to 90 min (with Doppler broadening) for problems involving 67 levels and unit lethargy widths. Input and theory are discussed, and a typical listing is given. (auth)
Date: May 1, 1963
Creator: Kelber, C.N.
open access

630A Maritime Nuclear Steam Generator: Status Report Number 1

Description: From foreword: The primary purpose of this document is to set forth the current status of the 630A Nuclear Steam Generator, under development for the U.S. AEC.
Date: September 12, 1963
Creator: General Electric Company. Nuclear Materials and Propulsion Operation.
open access

Absorption-Multistage Flash Distillation Process

Description: "The major factors which influence the cost of water production from sea water by distillation methods are (1) the cost of fuel or energy required by the distilling plant, and (2) the required capital investment. Preliminary studies on the application of absorption or solution cycles to distillation methods for saline water conversion indicated that the fuel cost or thermal economy of a distillation plant could be improved by combining the distillation process with an absorption or solution cycle" (p. 1).
Date: September 1963
Creator: Fluor-Singmaster & Breyer, Inc.
open access

Acceptance Test Facility Safeguards Report.

Description: The purpose of this report is to describe the operation of the Acceptance Test Facility (ATF) and testing of SNAP 10A Auxiliary Power Units (APU) in the facility.
Date: January 1, 1963
Creator: Soske, P. L.; Ostenso, A. S.; Kamensky, F. J. & Berger, S.
open access

Acid-Base Reactions in Fused Salts. Dichromate-Bromate Reaction

Description: Technical report. From Abstract : "The reaction of Lewis acid and base, Cr2O7= and BrO3_, in fused KNO3 - NaNO3 mixtures has been shown to involve an equilibrium followed by a slow decomposition to gaseous products."
Date: February 1963
Creator: Duke, F. R. & Schlegel, James
open access

Activation of electrical machinery. Supplement 1. [Preliminary evaluation; not applicable to ground tests]

Description: The following analysis of the induced radioactivity in SNAP-50/SPUR electrical machinery having a high cobalt content is submitted. Induced radioactivity in the flight vehicle will contribute negligibly to allowable radiation levels. This is especially so due to the low neutron to gamma ratio of assumed radiation damage tolerances to semiconductors. A calculation to estimate the order of magnitude of induced radioactivity in cobalt is attached. The calculation is based on a best guess of the neutron spectrum directly behind a lithium hydride shield. The resulting low cobalt activity and associated dose rate of about 1 mr/hr at 10 ft from a generator or a motor is insignificant. Although the evaluation indicates insignificant levels of induced radioactivity, this conclusion is not applicable to a ground test. Neutron moderation and scattering from a containment vessel and biological shield would greatly perturb the neutron environment behind the flight shield. Posttest handling of all components within the vacuum test chamber will undoubtedly be a problem. Notwithstanding the importance of limiting induced radioactivity, other considerations such as economy, cooling and vacuum requirements will largely dictate the final facility design. In summary, an activation analysis involves the overall facility design and will not be readily resolved. For a 10,000 hr. test the Co/sup 60/ activity may range from 100 curies per lb of cobalt where no shielding is provided to 10/sup -3/ curies per lb of cobalt where the equivalent of a flight shield is provided.
Date: November 15, 1963
Creator: Smolen, J.R.
open access

The Activity Coefficients of Hydrochloric Acid and Sodium Chloride in Hydrochloric Acid-Sodium Chloride Mixtures

Description: The activity coefficients of HCl and NaCl in HCl--NaCl mixtures were computed from literature data. The calculations are based on the observation that at constant ionic strength and temperature the logarithm of the activity coefficient of HCi in HCl--NaCl mixtures varies linearly with NaCl concentration. (auth)
Date: July 29, 1963
Creator: Lietzke, M. H. & Stoughton, R. W.
open access

ACTIVITY RELEASE FROM THE N.S. SAVANNAH IN THE MAXIMUM CREDIBLE ACCIDENT

Description: The release of fission products that would occur following the maximum credible accident aboard the N.S. Savannah has been examined. Four significantly different, but realistic, operating histories were considered. The rate of release of noble gases and of iodine isotopes as a function of time after the accident was determined for each operating history and for both normal and emergency reactor-compartment ventilation systems. The influence of radioactive decay and of the time delay in release and transport of activity through the containment system was investigated. Most of the results are expressed in terms of activity release and resultant individual exposures, although some consideration is given to population exposures and to the interpretation of these results in the light of stationary reactor site criteria. (auth)
Date: October 16, 1963
Creator: Anderson, T. D.; Buchanan, J. R.; Cottrell, W. B.; Fontana, M. H.; Klepper, O. H. & McCurdy, H. C.
Back to Top of Screen