Search Results

open access

Preliminary Design Study for a Sodium-Graphite-Reactor Irradiation Facility

Description: The results of an investigation to integrate a Na/sup 24/ irradiation processing facility with an operating sodium graphite reactor are presented. An irradiation facility incorporated into a reference SGR (Hallam Nuclear Power Facility, Hallam, Nebraska) is described. Development of the facility application, preliminary design criteria and capital and operating costs are discussed. Recommendations for further development of the technology and economics of this type of irradiation facility are i… more
Date: January 31, 1959
Creator: Thompson, D.S. & Benaroya, V.
Partner: UNT Libraries Government Documents Department
open access

DETERMINATION OF MAXIMUM PERMISSIBLE LEAKAGE FROM THE HRT PROCESS STEAM SYSTEM

Description: Calculations were made to determine the radiation hazard to HRT personnel as a result of leakage to the atmosphere from the process steam system in the event of a heat exchanger tube rupture. These calculations show that with the present four-minute delay before dumping approximately 1020 lb of fuel solution may be transferred to the steam system. The radiation hazard from fission products in the atomosphere will be negligble if the steam killer blower is operating. If this blower is not operat… more
Date: January 30, 1959
Creator: Gift, E.H.
Partner: UNT Libraries Government Documents Department
open access

Dissolution of Zirconium Matrix Fuels in Molten Fluoride Salts

Description: The application of the Volatility Process for reprocessing spent reactor fuels to zirconium fuel elements requires the conversion of the fuel to a fluoride form. This conversion can be accomplished by the dissolution of such fuels in a bath of molten fluoride salts with a hydrogen fluoride sparge. Studies with dummy zirconium fuel elements showed dissolution rates averaging 2 mg/sq cm-min. The effect of submergence, geometrical configuration, and HF impingement rate appeared more significant th… more
Date: January 30, 1959
Creator: Horton, R. W. & Whatley, M. E.
Partner: UNT Libraries Government Documents Department
open access

Quarterly Report of the Solution Materials Section for the Period Ending January 31, 1959

Description: Deposition of U, Cu, and Ni sulfates from dilute simulated HR fuel solution circulated through a heated Zircaloy-2 pipe section is reported. It was found that deposition occurred under boiling conditions in horizontal pipes because of a gas space which prevented wetting in the upper part of the pipe. Other salt deposition studies in an in-pile loop are reported. Effectiveness of oftwo different hydroclone mountings in removing corrosion products from a loop was studied. Results of chemical stab… more
Date: January 30, 1959
Creator: Griess, J. C.; Savage, H. C.; Greeley, R. S.; English, J. L.; Bolt, S. E.; Hess, D. N. et al.
Partner: UNT Libraries Government Documents Department
open access

Solid State Division Quarterly Progress Report: August 1952

Description: This quarterly progress report discusses the ongoing work within the Solid State Division at the Oak Ridge National Laboratory for the period ending August 10, 1952. Projects discussed include radiation metallurgy, engineering properties, fused salts, crystal physics, and solid state reactions.
Date: January 30, 1959
Creator: Billington, D. S. (Douglas S.) & Howe, J. T.
Partner: UNT Libraries Government Documents Department
open access

Neutron-Flux Measurements in a Concentric-Cylinder Fuel Element

Description: The following report presents neutron-flux measurements made with a concentric-cylinder element (Mark II) and includes axial, radial, and peripheral flux distributions.
Date: January 29, 1959
Creator: Anno, James N.; Fairand, Barry P. & Chastain, Joel W., Jr.
Partner: UNT Libraries Government Documents Department
open access

NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT

Description: Neutron-flux measurements in a concentric-cylinder fuel element were made in a gas-cooled in-pile loop operated adjacent to the core of the BRR. The fuel element comprised four concentric fuel cylinders. Each fuel annulus (outside diameters- 1.248 1.018, 0.810 and 0.590 in.) consisted of a 0.031-in.- thick core of UO/sub 2/ dispersed in type 347 stainless steel and clad on each side with 0.007 in. of typee 318 stainless steel. The element was 24 in. long and the total uranium-235 content was ap… more
Date: January 29, 1959
Creator: Anno, J.N.; Fairand, B.P. & Chastain, J.W. Jr.
Partner: UNT Libraries Government Documents Department
open access

The Decontamination of Reactor Cooling Water with Aluminum

Description: The discharge of cooling water from the Hanford reactors introduce radioactive contaminants to the Columbia River. These materials may subsequently bring about exposure to human populations either through the direct use of the water for sanitary purposes or transfers of the radioisotopes into the food chains. It is therefore desirable to keep to a minimum the amounts of radioisotopes released to the river.
Date: January 28, 1959
Creator: Silker, W. B.
Partner: UNT Libraries Government Documents Department
open access

A Simple Calibration and Checking Facility for Fast and Slow Neutron Detectors

Description: From introduction: The principal work described below is the development and tests of a slow neutron flux generator based on a generator developed at the University of California (UCRL 8359, W. Patterson, Roger Wallace, "A Method of Calibrating Slow Neutron Detectors"). It is shown that a tolerance flux density of slow neutrons results within a cubical cavity 15 inches on an edge with 4-inch-thick walls of water or paraffin wax when a plutonium-beryllium source emitting about 7 million neutrons… more
Date: January 28, 1959
Creator: Redmond, A. H.
Partner: UNT Libraries Government Documents Department
open access

STATUS REVIEW OF THE KEWB PROGRAM

Description: jectives, the accomplishments, and a summary of the work outstanding. The obtectives of the experimental and analytical studies were to investigate and reach an understanding of the kinetic behavior of aqueous homogeneous reactors. Information produced by the program, experiments on the spherical core, capsule experiments, and the remaining work schedule are discussed. (W.D.M.)
Date: January 28, 1959
Creator: Flora, J. W.
Partner: UNT Libraries Government Documents Department
open access

Waste Treatment and Disposal Problems of the Future Nuclear Power Industry

Description: The elements of waste treatment and disposal are assessed which are expected to become important in the development of the nuclear power industry of the future. Growth of the nuclear power economy is considered along with composition and quantities of anticipated waste. In addition, the economic implications of waste disposal are considered. It is concluded that research should be concentrated on decontaminating off-gases and on conversion of wastes to a more suitable form than liquid for stora… more
Date: January 28, 1959
Creator: Bruce, F.R.
Partner: UNT Libraries Government Documents Department
open access

Effect of saturation of water with dissolved corrosion product on the in-reactor corrosion rate of aluminum in deionized water: Proposal for test in 1706 KER

Description: This report discusses a proposal to measure the corrosion rate of aluminum-clad DOE elements in deionized water in one of the KER Loops at the highest obtainable surface temperatures under conditions where the bulk water adjacent to the elements is saturated with aluminum corrosion product. This will determine whether in-reactor corrosion may be reduced by controlling the aluminum concentration in the water and also provide corrosion data in deionized water at higher temperatures than is now av… more
Date: January 27, 1959
Creator: Dickinson, D. R.
Partner: UNT Libraries Government Documents Department
open access

Further Studies of Sintered Refractory Uranium Compounds

Description: The refractory uranium compounds UC and UC/sub 2/ were prepared by arc- melting methods, UN by gas solid reactions, and UB/sub 2/ and UBe/aub 13/ by solid-solid reactions. Techniques for sintering powder compacts of these compounds to densities 90% of theoretical or greater were developed. Mean linear thermal-expansion coefficients of 7.9 x 10/sup -6/ per deg F for Uc and 9.3 x 1O/ sup -6/ per deg F for UBe/sub 13/ were obtained in the range 68 to 1800 deg F. The thermal conductivity of sintere… more
Date: January 27, 1959
Creator: Tripler, A. B., Jr.; Snyder, M. J. & Duckworth, W. H.
Partner: UNT Libraries Government Documents Department
open access

NATURE OF RADIOACTIVE WASTES

Description: The integrated processes of nuclear industry are considered to define the nature of wastes. Processes for recovery and preparation of U and Th fuels produce wastes containing concentrated radioactive materials which present problems of confinement and dispersal. Fundamentals of waste treatment are considered from the standpoint of processes in which radioactive materials become a factor such as naturally occurring feed materials, fission products, and elements produced by parasitic neutron capt… more
Date: January 26, 1959
Creator: Culler, F.L. Jr.
Partner: UNT Libraries Government Documents Department
open access

SOME EFFECTS OF RADIATION ON SOLVENT EXTRACTION PROCESSES

Description: Presented at the 135th National Meeting of the A.C.S., Boston, April 5- 10, 1959. The yield of total acid, G/sub acid/, in the radiolysis of tributyl phosphate--Amsco 125-82 solutions is 2.7 times the electron fraction of TBP, or approximately 2.7 times the weight fraction of TBP, per 100 ev of energy absorbed by the solution. Dibutyl phosphoric acid, DBPA, constitutes about 85% of the acid. Radiolysis of TBP also results in the conversion of about 0.9 molecules of TBP to a polymer per 100 ev o… more
Date: January 26, 1959
Creator: Davis, Wallace, Jr. & Wagner, Robert M.
Partner: UNT Libraries Government Documents Department
open access

SOME EFFECTS OF RADIATION ON SOLVENT EXTRACTION PROCESSES

Description: The yield of total acid, G/sub acid/, in the radiolysis of tributyl phosphate-Amsco 125-82 solutions is 2.7 times the electron fraction of TBP, or approximately 2.7 times the weight fraction of TBP, per 100 ev of energy absorbed by the solution. Dibutyl phosphoric acid, DBPA, constitutes about 85% of the acid. Radiolysis of TBP also results in the conversion of abcut 0.9 molecules of TBP to a polymer per 100 ev of energy absorbed. Uranium extraction-stripping tests with an 8 stage spinner colum… more
Date: January 26, 1959
Creator: Davis, Wallace, Jr. & Wagner, Robert M.
Partner: UNT Libraries Government Documents Department
open access

Plutonium Recovery from Contaminated Materials Project CGC-813-Scope Revision No. 2

Description: An inventory of the contaminated materials accumulated since the initiation of this project in June 1958, revealed a larger variety and quantity of materials that could be burned, than was specified for the initial scope. Therefore, it is desirable to revise the scope to permit handling the majority of these materials with the initially installed equipment.
Date: January 23, 1959
Creator: Doud, E.
Partner: UNT Libraries Government Documents Department
open access

Scope design for conversion of Purex anion exchange prototype to a manufacturing facility

Description: This document is a HAPO report dated January 23, 1959, and describes the plutonium tail-end anion exchange system, installed in Purex as a prototype unit. Although some modifications, those considered most needed, were made to the unit, additional changes and refinements were still needed to convert the prototype to a fully acceptable manufacturing facility. This document covers the scope design of these modifications. The purpose of this document is to provide scope design criteria for a proje… more
Date: January 23, 1959
Creator: Gustafson, L. D.
Partner: UNT Libraries Government Documents Department
open access

Composition of Solids from Purex LWW

Description: The solids in Purex lww were first observed during flowsheet tests for recovery of fission products from plant wastes. Since the nature of this solid was not apparent from the flowsheet composition of lww, some work was performed to characterize this material. Although this work has been conducted over a period of about one year, it has been subordinate to the main one of testing flowsheets for fission product recovery. The solids have been observed in each of about six samples of plant lww tha… more
Date: January 22, 1959
Creator: Van Tuyl, H.H.
Partner: UNT Libraries Government Documents Department
open access

Control of Thorium Oxide Particle Size

Description: By carefully controlling the agitation, temperature, concentrations, and reagent addition rate, it was possible to produce a thorium oxide product within 0.3 mu of any selected size between 2.0 and 3.0 mu . The rate of addition of oxalic acid to a thorium nitrate solution was the variable used in the current series of tests. (auth)
Date: January 22, 1959
Creator: Johnsson, K. O. & Ellison, C. V.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen