Environmental Monitoring Semiannual Report: January-June 1962 (open access)

Environmental Monitoring Semiannual Report: January-June 1962

From summary: This report summarizes environmental monitoring results for the first six months of 1962.
Date: November 1962
Creator: Moore, J. D.; Fisher, W. L. & Rowe, P. A.
System: The UNT Digital Library
Multi-Channel Boiling Stability for Sodium Graphite Reactors (open access)

Multi-Channel Boiling Stability for Sodium Graphite Reactors

Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.
Date: March 1, 1962
Creator: Cappel, H. H.
System: The UNT Digital Library
Thermal Stress Testing of Beryllium Oxide Moderator Shapes (open access)

Thermal Stress Testing of Beryllium Oxide Moderator Shapes

Abstract: Perforated BeO plates were thermal shock tested to evaluate the effect of: (1) localized temperature variations adjacent to the perforations, and (2) radial gradients across the entire plate.
Date: August 1, 1962
Creator: Sujata, H. L.; King, M. & Waters, F. J.
System: The UNT Digital Library
The FAIM Code: a Multigroup, One-Dimensional Diffusion Equation Code (open access)

The FAIM Code: a Multigroup, One-Dimensional Diffusion Equation Code

Abstract: FAIM is a general multigroup, one-dimensional diffusion equation code programmed in FORTRAN language for the IBM 7090 computer.
Date: April 15, 1962
Creator: Baller, D. C.
System: The UNT Digital Library
Capsule Irradiation of Unalloyed Uranium at High Temperatures (open access)

Capsule Irradiation of Unalloyed Uranium at High Temperatures

Abstract: Cast and Wrought specimens and restrained wrought specimens of unalloyed uranium were irradiated in the Materials Testing Reactor, as the first in a series of experiments to develop fuel materials for sodium cooled reactors.
Date: August 15, 1962
Creator: Harrington, D. G. & Newton, J. B.
System: The UNT Digital Library
Evaluation of Zirconium Hydride as Moderator in Integral, Boiling Water-Superheat Reactors (open access)

Evaluation of Zirconium Hydride as Moderator in Integral, Boiling Water-Superheat Reactors

This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.
Date: March 1, 1962
Creator: Gylfe, J. D.; Rood, H.; Greenleaf, J.; Balkwill, K.; Prem, L. & Goldfisher, L.
System: The UNT Digital Library