Calcination of liquid radioactive wastes, the process of converting metal nitrates and sulfates to oxides by heat, is under development at Hanford as a means of reducing these liquids to a dry powder or solid which can be stored safely. Radiant-heat spray calcination, one of the methods under study, was first investigated at the Oak Ridge K-25 Plant (1) as a possible method of calcining uranyl nitrate to uranium trioxide. The process has also been under extensive development at the Pulp and Paper Research Institute of Canada (2) and is designated by them as the Atomized Suspension Technique.

It is recommended that self-concentration of Purex wastes should not be started until much more complete knowledge is obtained of the conditions and reactions in these wastes. The presently planned tests in the Semi-Works will be a practical demonstration of the problems but will be of little value on the initial Purex wastes as these tests are being run almost concurrent with the filling of the first Purex storage tank.

Thus we see that the derivation of elementary age theory is based on replacing the probability of collision governed by the Milne kernel by a Gaussian probability having the same zeroth and second moments. It is apparent that the age theory distribution will fail close to the source and at many free paths from the source.

Calculations are presented regarding the effect of delayed neutrons in reactors fueled with mixtures of Pu²³⁹ and U²³⁵. Results are plotted to illustrate the change in thermal flux as a function of time for various positive or negative step changes in reactivity.

aluminum has been considered as a material for cladding fuel elements to be used in reactors cooled by high-temperature recirculating water. Aluminum is abundant, chap, has a relatively low cross section for thermal neutrons, and is readily fabricated into complex shapes. For these reasons, it appears that the use of aluminum rather than steel, Zircaloy or other candidate materials for fuel element cladding, would result in considerable savings in operating expense. The actual economics will of course depend on specific factors such as core materials and design of the reactor/. Aluminum appears particularly attractive for cladding the cores of aluminum-plutonium alloy proposed for the Plutonium Recycle Test Reactor since, in this case, the physical characteristics of the core and cladding would be practically identical.

This document presents the bases for those PRTR Process Specifications involving considerations of heat transfer and fluid flow. It will be noted that some of the bases incorporate information gained from laboratory experiments with electrically heated test sections simulating PRTR fuel elements. These experiments are described in some detail elsewhere.

From introduction: "Five appendices, each of which pertains to a separate phase of the Filtration of Radioactive Aerosols by Glass Fibers program."

Report describing a study regarding the solubility and related properties of tributyl phosphate (TBP), which has lower solubility in water than most uranium solvents. Pure TBP and TBP in an inert diluent were tested for solubility in water, several nitric acid solutions of varying concentrations, and other solutions typical of process streams. For TBP-diluent-water system tests, equilibrium water content in organic phase was found.

A previous study has been completed on the buildup and decay of xenon poisoning in the PRTR by simulating the reactor as a single region. The results of this study indicated that a study using a more refined model of the reactor would be valuable.

The purpose of this work was to produce a reliable, inexpensive instrument for the detection of hard beta contamination on the clothing of personnel. Also, the monitoring was required to be accomplished without interrupting the normal activities of the employees.

A correlation between the optical densities if the Pb-Sn alloy system and UO2 with respect to Co^60 gamma radiation has been determined. This enables one to fabricate density standards of whatever geometry may be desired for one in the gamma absorptiometer by simply casting a Pb-Sn alloy of the proper composition to correspond to the density required.

Objective: "The observations summarized in the present report were made to determine the stability of ferrous and sulfamate ions and sulfamic acid in RA Column scrub solution (RAS) and in aqueous solution comparable to the RA Column feed tee composition (RAFS). Knowledge of the stability of the reducing agent (ferrous iron) is necessary to properly plan the method of preparation and introduction of this component into the RA system and to assure dependable and reproducible performance."

Buckling measurements were taken with hollow natural uranium slugs in graphite-moderated lattices, in which the lattice spacing was varied from 6 3/16 to 15 inches in five steps. A 1.66" O.D. slug was used with two different core sizes, 1.11" I.D. and 0.81" I.D. The results of wet (water in the core) and dry measurements are reported. Buckling curves are given showing the effect on the buckling of coring the slug and placing water in the core. Simple three-region diffusion theory is applied to the measured lattices in an effort to correlate measurement with theory, but with only limited success.

The primary purpose of a cleanup system in a recirculating water loop is to maintain the best possible water quality conditions. This is normally accomplished by continuously purifying all or a portion of the coolant. A secondary objective of the cleanup system is to help maintain the system pH at a constant value. A system that will satisfactorily accomplish both of these objectives is at times difficult to obtain. Generally the pH control characteristics are sacrificed in favor of the more important cleanup requirements. A somewhat new approach to the problem pf cleanup system design appears to offer a solution to this problem for high pH systems.

Facilities shall be provided in the 221-U Building for storing the metal solution product of the dissolution step in existing tankage from U, T, and B Plants until a reprocessing campaign is scheduled through Redox. This section shall provide a sampling tank for fuel accountability sampling and a pump tank from which the solution will be pumped via a cross-country pipeline to Redox for further processing.

At the request of 234-5 Development, studies of a system for controlling plutonium powder transfer were undertaken by Process Equipment Development. This report presents details of equipment designed to fulfill requirements and proposes equipment for installation on additional Hot Button Line prototypes.

A series of tests conducted in an attempt to reduce the observed O2 content of primary recirculation loop water to a value of <0.14ppm by the current loop method which is a combination of pressurizer degasification and venting flow.

The ambient temperature of the Columbia River is of interest to the Aquatic Biology Operation since it provides a basis for controlling the temperatures in various experimental tanks which contain Columbia River organisms and since it provides some indication of whether conditions are favorable for the valuable species of fish living in the river. Since the start-up of the plant, temperatures of the river water have been taken in the several water treatment plants.

This paper discusses technical bases, operating standards, and instrumentation systems required to achieve the goal of nuclear safety in operation of high-level reactors. Because of the relatively long neutron lifetime in a thermal reactor and the comparatively modest time rates of reactivity change associated with fission poison and temperature effects, the rod system used for compensating reactivity transients and for maintaining flux distribution control may be operated manually. Manual operation of the reactor and the prediction of pile reactivity statue during outages of course involve factors of human error, normal reaction time, and judgment. It is the aim to specify procedures for the operation such that manual control of the reactor may be adequately maintained at all times within the range of reasonably expected calculational errors and human reaction times. The technical considerations used in developing these standards, including the available monitoring indications are discussed first; then the automatic trip devices which should back up the procedures are also mentioned.

Limitations on the maximum temperature of the pile moderator have been imposed to prevent or control the gasification reaction of the graphite stack with the pile gas atmosphere. An uncontrolled reaction would cause a loss of structural strength, density, and thermal conductivity and would thus result in an unnecessary shortening of the effective operating life of the HAPO piles. Numerous studies have been made to define the rates of reaction of the pile stack with various components of the pile gas and the influence on these rates of various pile operating conditions. Based on these studies, maximum graphite temperature has been so restricted that after 10 years of pile operation the graphic stack has sustained less than 1% oxidation (based on recent experimental data). As pile power levels increased, the maximum graphite temperature has also increased and will increase still more as present and future power level programs are completed. In order to evaluate the effect on the graphite stack of the present programs, and to some extent determine the feasibility of future programs, a more realistic approach to the question of structural integrity and oxidation versus pile life is necessary. It is the purpose of this report to summarize …

Large savings in waste storage space may be realized by lowering the pH at which TBP waste is stored. Additional savings in neutralizing chemicals and operating time would also increase the monetary gain from such a process change. However, before such a change could be made, the corrosive effect of TBP waste at a lower pH on the mild steel waste storage tanks had to be determined.

Basic Studies. It has been reported previously that mixed crystals of PuO2 and UO2 have a higher sintering rate than UO2 alone. However, results to the contrary were obtained on addition to 1/4, 1/2, 1, 2, 5, and 10 w/o PuO2 to PWR grade UO2. Pellets of the above concentrations were heated for one hour in hydrogen at 100 C intervals from 1000 to 1600C.

Report discussing the use of a small-source approximation for engineering calculation of thermal piles. Explanation for how this type of approximation is more appropriate in certain problematic situations where a homogenization approximation would be doubtful is given.

Purex process which utilizes solvent extraction to separate and purify uranium and plutonium from each other and from fission products contained in irradiated uranium fuel elements.