From introduction: "The over-all objective of the present study was to determine the nature and extent of the present study was to determine the nature and extent of the corrosion of these materials under a variety of such storage conditions and to determine the ability of several potential protective coatings to retard this corrosion."

From introduction: "This article gives an account of the determination of the complete crystal structures of the normal orthophosphates of barium and strontium."

Abstract: "Theoretical equations have been derived for calculating heat transfer coefficients for fluids flowing through concentric annuli for the following two cases: (A) constant and equal heat fluxes from both walls, and (B) constant, but unequal, heat fluxes from the walls, with equal wall temperatures at a given axial position along annular channel. In the derivations, the conditions of fully-established flow, and independence of physical properties with temperature variation across the flow channel, were assumed. The only geometrical parameter in this general case is the radius ratio r2/r1, and in the study it was varied from 1.0 to 10.0."

In conjunction with the fuel element development program at Hanford, it is desired to determine the effects of cladding and core temperatures, cladding thickness, and exposure upon the swelling behavior of unalloyed uranium. To obtain this information, it is proposed to irradiate several fuel rods, clad by coextrusion with Zr-2, in NeK filled stainless steel capsules. The central uranium temperatures are to be monitored by axial thermocouples. Irradiation tests in the MTR and ETR using capsules of similar design have been and are now being conducted. GKH 3-31, a fuel rod, clad with 0.030" Zr-2, operated in the MTR at an average center fuel temperature of 425 C to an exposure of 2100 MWD/T. GKH 3-57, 3-58, and 3-59 are presently undergoing irradiation in the ETR.

In order to obtain the most economical utilization of underground storage facilities it is desirable to maintain a running inventory of heat generation and available self concentration in a given tank. Further, it is believed that such knowledge will be helpful in studying underground storage technology. The calculation of fission product heat generation and available self concentration factor in separations waste storage tanks is a complex process. The complexity is increased greatly when material of varying irradiation history, cooling time, etc., is stored at varying production rates. This document presents in tabular form the power generated from the fission product activity associated with one ton of irradiated uranium using the various operating conditions and decay periods of interest for waste storage considerations.

"A small quantity of a Np-237, the long-lived ancestor of the 4n + 1 series, has been isolated from a natural source. The Th-229 content of the Th-230 obtained from Belgian Congo pitchblende concentrate and of Th-232 obtained from Brazilian monazite has been determined indirectly through isolation of the Ac-225 daughter. The mass ratio of Np-237 to U-238 in the Belgian Congo pitchblende concentrate has been determined as (1.8 +/- 0.4) x 10-(-12). The mass ratio of U-233 to U-238 in the same ore deposit is found to be (1.3 +/- 0.2) x 10-(-13). The ratio of the respective neutron-capture cross sections of U-238 in the ore body for the (n, 2n) and (n, gamma) reactions is calculated to be (1.3 +/- 0.6) x 10-(-3) on the basis of the relative Np-237 and Pu-239 contents."

The purpose of this report is to summarize the work done at KAPL on physical properties of high density concretes.

This twenty-second quarterly report of the research and development activities of the Radiological Sciences Department, Hanford Atomic Products Operations, as before, includes some items charged to control but included for general interest. Such are identified as "not charged to research." The previous report of this series was numbered: HW-34408 January 10, 1955

Crib evaluation studies showed the soil uptake of plutonium from the Recuplex process (CAW) waste to be low. Preliminary studies of the low soil adsorption of plutonium revealed the presence of a plutonium nitrate anion complex which could be removed by adsorption on a strong base anion exchange resin.

Thermal Resistance of the contact between to core and the jackets or unbonded fuel elements may easily be the largest source of error in core temperature predictions. The object of this work is to improve contact conductance predictions by measuring conductance of the joint between reactor fuel and cladding materials at joint pressures, temperatures and thermal flux levels approaching reactor service conditions.

A falling ball viscometer under development at ORNL, employing a flow system and an electromagnetically operated dash-pot pump, was evaluated for possible use with aqueous ThO2 slurry systems under reactor irradiation. Interchangeable check valve inserts were designed and fabricated to investigate several pump designs. Magnetic flux concentrations which originally prevented circulation of the ferritic stainless steel viscometer ball through the electromagnetic pump were eliminated by substitution of iron-magnetic stainless steel inserts. Viscosity was correlated through a logarithmic plot of the dimensionless Reynolds number vs. Froude number with the ratio of ball diameter to tube diameter as a parameter. The relation is linear in the laminar flow region.

The following report discusses the results of a survey that was initiated to ascertain the availability of uranium hydride in tonnage quantities.