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Semipermanent Freeze Plug Tests for HRT-CP (open access)

Semipermanent Freeze Plug Tests for HRT-CP

Five uninsulated semipermanent freeze coils, series connected, can be frozen when submerged in 70 F water with a Freon-11 flow rate of 1.85 gpm at an inlet temperature of -40 F. The refrigeration unit of the HRT-CP is capable of delivering >3 gpm to a similar semipermanent freeze coil system located in Cell C. Therefore the number of F-11 risers required in Cell C of the HRT-CP can be minimized by series connecting this many semipermanent freeze coils where required.
Date: March 11, 1957
Creator: Winget, R. H.
System: The UNT Digital Library
Neutron and Gamma-Ray Attenuation for a Fission Source in Water : Comparison of Theory with LTSF Measurements (open access)

Neutron and Gamma-Ray Attenuation for a Fission Source in Water : Comparison of Theory with LTSF Measurements

Calculations are presented of the fast neutron dose rate, the gamma-ray dose rate, and the thermal neutron flux along the source plate axis in the Lid Tank with the tank filled entirely with water. These calculations are compared with experimental measurements. The calculations are absolute and are compared with absolute measurements.
Date: March 12, 1957
Creator: Otis, David Ridgway
System: The UNT Digital Library
In-Pile Slurry Loop Program (open access)

In-Pile Slurry Loop Program

The in-plie slurry loop work is now being considered as a joint program between the PAR project and ORNL. It is proposed that PAR design, fabricate and test the in-pile loops and that ORNL operate the loops in-pile, dismantle the loops after irradiation and made the appropriate measurements to determine the radiation effects. This report gives the objects of the slurry in-pile program and outlines the facilities and operations required to execute ORNL's part of this program.
Date: March 8, 1957
Creator: Arehart, T. A.; Compere, E. L. (Edgar L.); Ferguson, D. E.; Korsmeyer, R. B. & McBride, J. P.
System: The UNT Digital Library
High-Temperature Properties in Relation to Design (open access)

High-Temperature Properties in Relation to Design

The essential physical and mechanical properties of materials which enter into the design considerations of high-temperature systems have been outlined, and the relationship of these factors to the detailed design of structural members noted. Particular attention is given to the phenomenon of strain-cycling and relaxation, and several engineering devices wherein these effects can markedly influence design are mentioned. Techniques and apparatus for the measurement of strain-cycling and relaxation properties are described and some recent data for Inconel at 1500 F is reported. A brief discussion is presented on design-criterial for high-temperature applications, and it is pointed out that in some high performance devices the considerations of the strain-cycling phenomenon may impose the design philosophy of "steady-state" operation. Some elementary applications of strain cycling data to design are described, and test experience indicates that analytical techniques can be developed for predicting with reasonable reliability of the fatigue life of structures.
Date: March 14, 1957
Creator: Weir, J. R.; Meghreblian, Robert V. (Robert Vartan) & Douglas, D. A.
System: The UNT Digital Library
Comments on the Determination of the Particle Size Distribution of Thorium Oxide (open access)

Comments on the Determination of the Particle Size Distribution of Thorium Oxide

Factors affecting the results of thoria particle size distribution measurements by sedimentation procedures currently and recently employed are considered. The effects of thoria concentration, solvent, dispersant, thoria properties, and other factors are discussed.
Date: March 31, 1957
Creator: Moore, G. E.
System: The UNT Digital Library
TBP Stripping in Bubble-Cap Column and Concomitant Product Evaporation (open access)

TBP Stripping in Bubble-Cap Column and Concomitant Product Evaporation

A study was conducted to demonstrate the stripping and evaporation steps in a Purex-type uranium recovery process and to ascertain the operating behavior of the equipment under a range of conditions. The factors considered were control, effectiveness, and reliability of equipment and optimum feed point. Experimental procedures are described, and recommendations for equipment modifications are included. (J.R.D.)
Date: March 18, 1957
Creator: Long, J. T.
System: The UNT Digital Library
Beryllium (open access)

Beryllium

The information concerning beryllium assembled herein has been selected primarily to provide information that might be pertinent to the design of the ART. The creep-rupture data in the literature have been obtained from vacuum cast and extruded beryllium, rather than from the hot-pressed beryllium which will be used in the ART. A research program is being conducted by the Brush Beryllium Company, under contract to the Oak Ridge National Laboratory, which will eventually provide the design engineers with data on the high temperature strength properties of hot-pressed beryllium. In the interim, it will be necessary for the metallurgists and design engineers to make judicious estimates and extrapolations from the data available.
Date: March 21, 1957
Creator: Whitman, M. J.
System: The UNT Digital Library
Measurement of the Friction Characteristics for Flow in the ART Fuel-to-NaK Heat Exchanger (open access)

Measurement of the Friction Characteristics for Flow in the ART Fuel-to-NaK Heat Exchanger

The friction characteristics of a full-scale straight-tube model of the ART fuel-to-NaK heat exchanger were determined experimentally. The presence of the spacers resulted in a transition to semi-turbulent flow at a Reynolds modulus of 350. this semi-turbulent flow persisted up to a Reynolds modulus of 5,000. The circumferential spacers were found to contribute slightly more than the radial spacers to the pressure loss in the heat exchanger.
Date: March 19, 1957
Creator: Cohen, S. I. & Jones, T. N.
System: The UNT Digital Library
In-Reactor Autoclave Corrosion Studies. II, Autoclave Z18 (open access)

In-Reactor Autoclave Corrosion Studies. II, Autoclave Z18

In order to test the effect of chromate ion on the corrosion of Zircaloy-2, a 0.04 m uranyl sulfate solution (enriched) containing added acid, copper sulfate and 557 ppm of Cr (VI) was autoclaved with rocking at 280 degrees C for about eight days in the HB-5 facility of the LITR. The average corrosion rate established by the rate of consumption of oxygen was 9.9 mpy at a power density of 4.9 w/ml. It is indicated by comparison with a previous corrosion study under LITR radiation that the presence of Cr (VI) had no significant effect on the radiation corrosion of Zircaloy-2 by enriched uranyl sulfate solutions. However, the data are not conclusive and may be interpreted as showing a low corrosion rate for a limited period (i.e. a short term inhibition) followed by correspondingly rapid corrosion.
Date: March 22, 1957
Creator: Warren, K. S.; Davis, R. J. & Jenks, G. H. (Glenn Herbert), 1916-
System: The UNT Digital Library
Status Report on the Disposal of Radioactive Wastes (open access)

Status Report on the Disposal of Radioactive Wastes

The new and as yet unsolved problems introduced by the production of large quantities of fission products and radioactive isotopes from fission or neutron capture present mankind a most complex technical, economic, and political problem. On one hand, the possibility of using the fission process to produce energy from an unexploited and abundant natural source is emerging from large programs of research and development. We are also beginning to see the promise of use of particulate and electromagnetic radiation for the good of man. On the other hand, we are presented with the problem of controlling the dangerous products of fission for periods of time measured in terms of many hundreds of years, periods longer than the effective tenure of any political state in history. We must not only devise ways of protecting ourselves in the present and for our lifetime but, in addition, we must establish the basic technical, social, and administrative control of vast quantities of artificial radioactivity that must remain effective for at least ten to twenty lifetimes.
Date: June 25, 1957
Creator: Culler, Floyd L., Jr. & McLain, Stuart
System: The UNT Digital Library
Section 9.0 to Status Report on the Disposal of Radioactive Wastes (open access)

Section 9.0 to Status Report on the Disposal of Radioactive Wastes

Section 9.0 is the "Chemical Processes for Fission Product Concentration, Removal or Fixation" section of the Status Report on the Disposal of Radioactive Wastes. The report is divided into four areas: (1) Introduction; (2) Summary of waste processes; (3) Concept of a multipurpose waste processing facility; and (4) Details of some of the waste processes.
Date: September 3, 1957
Creator: Culler, Floyd L., Jr.
System: The UNT Digital Library
Use of Convergent and Asymptotic Series for Computation of the Modified Bessel Function of the First Kind. (open access)

Use of Convergent and Asymptotic Series for Computation of the Modified Bessel Function of the First Kind.

In the Communications of the ACM of April, 1960, p. 240, there appeared two Algol algorithms for the computation of In, the modified Bessel function of the first kind. One of these algorithms uses a convergent series and the other an asymptotic series. Their author, Dorothea S. Clarke, did not give any information with regard to which algorithm is more appropriate in a given situation (a given situation meaning the computation of In (x) for a definite pair of numbers n and x). Such information is necessary, however, if the algorithms are to be useful, particularly since the asymptotic series is in many cases not valid.
Date: December 12, 1960
Creator: Bumgarner, L. L.
System: The UNT Digital Library
Analog Computer Analysis of the Thermal Behavior of the EDCR Due to Reactivity Accidents (open access)

Analog Computer Analysis of the Thermal Behavior of the EDCR Due to Reactivity Accidents

The ORNL analog computer was used to examine one aspect of a safety system for the EGCR. The thermal behavior of the fuel as a result of increases in reactivity at various power levels and gas flows was examined. Safety actions initiated by wither a flux-minus-flow signal or a reactor outlet gas temperature signal were included in simulation.
Date: December 16, 1960
Creator: Pierce, R. M.
System: The UNT Digital Library
Steady-State Radial and Axial Temperature Profiles of an E. G. C. R. 5 1/2" O. D. Through Tube During 1500 KW Operation - Part II (open access)

Steady-State Radial and Axial Temperature Profiles of an E. G. C. R. 5 1/2" O. D. Through Tube During 1500 KW Operation - Part II

The steady-state radial and axial temperature profiles of an E. G. C. R. 5 1/2" O. D. through tube are determined for the test condition of an attemperated fuel assembly operating at 1500 KW in the loop. The profiles are determined for the case of the central control rod fully inserted and bank insertion to 62 inches (Δk = 0.025). The data are presented in the form of tables and curves.
Date: December 30, 1960
Creator: Noggle, L. W.
System: The UNT Digital Library
Thermal-Neutron Flux at the HRE-2 Core Wall, Determined from Zircaloy-2 Induced-Activity Measurements During Runs 13, 14, 16, 17, 18, 19, 20 and 21. (open access)

Thermal-Neutron Flux at the HRE-2 Core Wall, Determined from Zircaloy-2 Induced-Activity Measurements During Runs 13, 14, 16, 17, 18, 19, 20 and 21.

The thermal-neutron flux prevailing near the upper hole in the HRE-2 core wall was determined from measurements of the induced activity in a specimen cut from the edge of the hole. The specimen was removed following run 21 and was thus exposed to reactor neutrons during runs 13, 14, 16,17, 18. 19, 20, and 21. The neutron flux, calculated for an average core power level of 2 Mw, was 2.8 x10^13 neutrons/cm^2 sec. the core wall area from which the specimen was removed was located approximately 16-in. from the core center.
Date: December 20, 1960
Creator: Baker, J. E.
System: The UNT Digital Library
Homogeneous Molten Salt Reactors (open access)

Homogeneous Molten Salt Reactors

Multigroup one-dimensional calculations were done recently to obtain estimates of critical masses, power density distributions and fissioning spectra for some homogeneous molten salt reactors having outer reflectors and central "islands," placed inside the currently proposed MSRE vessel. For a 5-inch-thick outer reflector and 1-ft-diamter island, both beryllium, the calculated critical mass is 108 kg; 40 percent of the fissions occur at thermal, and the maximum power density of 3.9 times the core mean power density occurs at the island-salt interface. If the reflector thickness is increased to 10 inches, the critical mass is reduced to 34 kg; 67 percent of the fissions occur at thermal, and the peak power density of twice the core mean again occurs at the core island-salt interface.
Date: December 13, 1960
Creator: Nestor, C. W., Jr
System: The UNT Digital Library
Status Report on the Pneumatic Temperature Measuring system Studies for the EGCR (open access)

Status Report on the Pneumatic Temperature Measuring system Studies for the EGCR

The pneumatic temperature probe (PIM) is a device for measuring gas temperatures by utilizing the dependency of the flow of gases through a restriction on the temperature and pressure conditions. The determination is made by measuring critical mass flow across a restrictive element such as a nozzle and by knowing the upstream pressure and other variables pertinent to the critical flow equation, computing the temperature at the entrance to the restrictive element. In practice it has been found useful to use two critical flow nozzles in series and measure the ratio of the pressures at the nozzle inlets, together with the temperature at the downstream nozzle. The limitations of present thermocouple materials for long term use at elevated temperatures makes resort to this device attractive on many installations.
Date: December 30, 1960
Creator: Hochreiter, H. M.
System: The UNT Digital Library
After Shutdown Heating in the HFIR (open access)

After Shutdown Heating in the HFIR

After shutdown heating rates have been calculated for the target, control plates, and the beryllium reflector of the HFIR. Hilvety previously calculated the after shutdown heating rates in the element and his results are reported in ORNL CP-60-4-110. Decay times of 1.0 10, 10^2, 10^3, 10^4, and 10^5 seconds have been considered, and heat fluxes have been computed for all of the mentioned components. The greatest heat fluxes were found to be at the surfaces of the control plates and the permanent beryllium reflector.
Date: December 29, 1960
Creator: McLain, H. A.
System: The UNT Digital Library
Source of Fission Products in LITR Cooling Water (open access)

Source of Fission Products in LITR Cooling Water

A study has been made of the radioactive nuclides present in the cooling water of the Low Intensity Test Reactor. Prominent nuclides which have been identified include Np239 and certain fission products. The experimental evidence points to the probability that there could be adequate U235 contamination on the fuel plates to quantitatively account for the fission products identified in the LITR cooling water. inasmuch as new aluminum does not contain uranium, it must be assumed that the contamination occurs during the fabrication of the fuel elements.
Date: March 26, 1957
Creator: Moeller, D. W. (Dade W.) & Leddicotte, G. W.
System: The UNT Digital Library
Texas Board of Examiners of Psychologists Annual Financial Report: 2018 (open access)

Texas Board of Examiners of Psychologists Annual Financial Report: 2018

Annual financial report of the Board of Examiners of Psychologists documenting income, expenditures, and other relevant financial information for fiscal year 2018.
Date: September 21, 2018
Creator: Texas State Board of Examiners of Psychologists
System: The Portal to Texas History
Texas State Lunatic Asylum Annual Report: 1912 (open access)

Texas State Lunatic Asylum Annual Report: 1912

Annual report of the Texas State Lunatic Asylum, providing data about operations, updates about programs, and other information related to activities at the institution for the fiscal year 1912.
Date: 1912
Creator: Texas State Lunatic Asylum (Austin, Tex.)
System: The Portal to Texas History
Texas Panhandle Centers [Agency Newsletter], Volume 19, Number 2, February 2022 (open access)

Texas Panhandle Centers [Agency Newsletter], Volume 19, Number 2, February 2022

Monthly newsletter discussing news and events related to the Texas Panhandle Centers Behavioral and Development Health and other information about mental health.
Date: February 2022
Creator: Texas Panhandle Centers Behavioral and Development Health
System: The Portal to Texas History
Tests of Models of Nuclear Reactor Elements (open access)

Tests of Models of Nuclear Reactor Elements

"These two parts were issued separately, but are cataloged as a unit. For a model of a blanket sub-assembly of the proposed Atomic Power Development Associates reactor, laboratory measurements have been made of the head losses for the blanket element and for the various distributing orifices. Tests with water have been generalized so that the results provide design criteria for liquid sodium which is to be used in the reactor."
Date: March 1957
Creator: McNown, J. S.; Yih, C. S.; Yagle, R. A. & O'Dell, W. W.
System: The UNT Digital Library
Tests of Models of Nuclear Reactor Elements (open access)

Tests of Models of Nuclear Reactor Elements

"For a model of a blanket sub-assembly of the proposed AODA reactor, laboratory measurements have been made of the head losses for the blanket element end for the results provide design criteria for liquid sodium which is to be used in the reactor."
Date: March 1956
Creator: McNown, J. S.; Yagle, H. A. & O'Dell, W. W.
System: The UNT Digital Library