From preface: The following bibliography on high pressure-high temperature, and high pressure, is the result of an investigation into the equipment and experimental technique needed to increase the density of powdered compounds, especially those compounds which possess a high vapor pressure of hydrogen at moderate temperature and pressure.

The following report records a total of forty-nine graphite core samples that have been bored from process tube blocks in the Hanford piles with the intention to provide results to indicate that core boring is an entirely practical operation over bore scraping in obtaining powder samples.

The following report calculates pile reactivity changes that result from different pile loadings where it is desirable to have available values of (flux)2 d(volume) for various flat zone radii. This information was found for a cylindrical pile having circular flat zones and an effective radius of 598 cm., by integrating over a cross sectional volume of one cm. thickness.

Objective: The objective of this study is to develop a simple method of either removing the C1- ion from the RAW or converting the ion into a less corrosive valence state.

Abstract: "Nuclear emulsion has been made in a manner similar to that of Demers and of Perfilov. An attempt has been made to achieve the grain size and sensitivity of Perflov's type P-9 Sensitive emulsion. Silver bromide grain sizes of essentially the same average size as Perfilov's were made from precipitations carried out over a narrow range of pAg's. Three brands of gelatin were tried. High sensitivity has not been achieved."

This report summarizes progress on investigation into improving the technological limits of boiling water reactors, Vallecitos Boiling Water Reactor and other facilities.

A series of diffusion length measurements were carried out on graphite stacks of various constructions in an attempt to resolve the discrepancies between the graphite diffusion lengths measured in the Hanford reactors and that value measured in the Hanford Standard Pile. It was found that the diffusion length of the graphite in the Hanford reactors is in good agreement with the value for the Hanford Standard Pile when corrections are made for the absorption and scattering of neutrons by the aluminum process tubes, and for neutrons streaming in the channels of the reactors.

A method for determining the concentration of Li6 in aqueous solution has been tested using the nuclear reactions Li6 (n,α)H3 and O16 (H3,n)F18. Annihilation 7 radiation of induced 1.87 hour F18 radioactivity was counted with a well-type scintillation counter, and the radioactivity per millimole of lithium was found to be independent of lithium concentration below about 0.2moles/liter. The sensitivity limit for detecting lithium is less than 0.1 micromole (0.0075 micromole Li6).

The following document provides figures with all information that describe the heavy water moderated chain reacting pile at Argonne National Laboratory.

In a previous paper a method of measuring the equilibrium constants at various temperatures for the titular reaction has been described. This report applies the method to an investigation of the corresponding reaction of americium, representing a part of a systematic program of investigation of the thermodynamic properties of analogous reactions of lanthanide and actinide elements.

"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 following report discusses the building design and development of the reactor building, Building 1105, for Argonne National Laboratory, that is meant to house the reactor.

The following report examines a set of beryllium metal rods through pile radiation, taking measurements, and analyzing differences between irradiated and radiated rod samples.

The following report records a total of forty-nine graphite core samples that have been bored from process tube blocks in the Hanford piles with the intention to provide results to indicate that core boring is an entirely practical operation over bore scraping in obtaining powder samples.

The following report provides tables that measure the pressure of hydrogen fluoride at given temperatures and the hydrolysis of uranium tetrafluoride.

Abstract: "A model is presented for radiation damage to ductile metals that accounts qualitatively for the effects of radiation on the electrical and mechanical properties. The model consists of an agglomeration of interstitially displaced atoms into stacking faults in the lattice."

Writing the VARI-II Program was motivated by the need for a method of analyzing the results for the Absorber Burn-Up Experiment in progress at the Vallecitos Atomic Laboratory.

Ultrasonic equipment has been developed for nondestructive testing of Hanford fuel elements. The ultrasonic method has replaced the Frost Test for bonding layer inspection in the Hanford canning line, and provides more accurate and reliable results at lower cost. The method has also been adopted to the testing of new fuel elements for which no other method is available.

The purpose of this study was to develop equations for calculating fision product heating in the HRT-CP underflow pot from measured temperatures and to attempt to correlate the rat of solids accumulation in the underflow pot with fission heating and reactor power. Using fission heating data calculated from relating solids accumulation and heating have been tested. In one case an error of no greater than 26% was incurred in the calculation of the total weight of solids collected during chemical plant runs 17-4, 17-5, and 17-6. Further development work will be done on this correlation.

A research and development program aimed at devising processes for the economical recovery of the potentially valuable long-lived fission products from Purex waste has been under wat at Hanford for several years. When this work has begun, the concentrated waste was primarily a nitric acid solution (6 to 10 M HNO3) containing the fission products and relatively small concentrations of iron, sulfate, and other corrosion products. Flowsheets based on classical separation schemes and rather similar to processes used by the Isotopes Division at the AEC's Oak Ridge operation served to separate the desired fission products from one another and from the corrosion products (1,2,3).These separation schemes employed careful step-wise pH adjustment to precipitate first the iron and then to separate the desired fission products from one another. The flowsheets were demonstrated on a pilot-plant scale with full-level plant waste. However, since the earlier work was complete, plant operations have been modified....

Considerable interest has developed in the use of ion-exchange in the nuclear energy field in the last decade. Aside from the obvious use of providing demineralized coolant water for reactors, the projected uses of ion-exchange include the recovery of fission products from aquaeous waste streams and the separation and purification of fissionable materials from spent reactor fuels. The latter process may be incidental to the over-all operation, as is the case with the Purex anion exchange facility, or it may be the prime separation process, as may be the case in the recovery of Pu or U from spent power reactor (PRTR) fuel elements.

Thermal oxidation of graphite in flowing CO2 is being studied at 650 to 850 C, in a single-pass gas system at atmospheric pressure, by observing weight loss rates. The method is used to provide comparative data for candidate reactor graphites. The effects on oxidation rates of graphite purity, structure, coke type, graphitization temperatures and other manufacturing variables are determined. In addition, the effects of gas flow rates and graphite surface to volume ratios are observed.

A knowledge of the mechanical and physical properties of uranium is important in understanding and explaining its pile behavior. The effects of irradiation on the properties of uranium will be dealt with in another chapter. However, it is important to know and understand the pre-irradiation properties of uranium prior to investigating the effects of irradiation.

Four capsules containing Al-1.65 w/o Pu and Al-12 w/o Si-1.65 w/o Pu were charged into the MTR. These capsules will be irradiated to a burnup of 80-100% of the plutonium atoms to determine the stability of the material at high exposures. An additional sixteen capsules containing 5 to 20 w/o Pu in Al and Al-Si have been prepared. Eight are awaiting reactor space and should be charged soon. The remaining eight are being recanned to increase the end gap between the core and the can. Also four capsules containing high density UO2-PuO2 pellets canned in Zircaloy are awaiting assignment of reactor space.