Abstract: "The critical mass both of an untamped and water-tamped sphere with parabolic radial variation of the intrinsic buckling is calculated by a perturbation method. The result is applied to finding the minimum critical mass of plutonium in water suspension with infinite water temper; the calculations show that the minimum critical mass at constant concentration may be reduced by an amount of the order of 7.6% when the concentration of plutonium is permitted to vary throughout the suspension."

Report describing the temperature in the Brookhaven concrete shield proposed for the X reactor. The report details the methods and equations used to find the values that determined that the effect of the gun barrel assembly in cooling the concrete is negligible.

From abstract: "The present paper proposes a lattice test pile, (or LTP) as an alternate to the continuing program of building up and tearing down of exponential piles."

Report discussing a study regarding the identification and property analysis of plutonium (IV) monobutyl phosphate, which formed as a white precipitate in another study of the effects of monobutyl phosphate on plutonium(IV) in tributyl phosphate extractions.

Report discussing "the effect on the multiplication constant of using graphite of various density." Results are tabulated and plotted.

Introduction: "Evidence of masonite deterioration in test hole plugs was interpreted to mean that a previous estimate of the minimum lifetime of the masonite in a Hanford biological shield might be somewhat optimistic. The importance of understanding the effects of radiations on masonite for both present and future pile shields dictated that masonite be sampled from an operating pile shield and studied. The results and techniques are discussed in the following sections of this report."

The objective of this investigation was to determine the feasibility of centrifugal casting to produce fuel element plates suitable for use in a power reactor. Preliminary investigations involving the centrifugal casting of metal plates suitable for use in a power reactor proved to be successful. Lead plates cast at medium speeds contained very few voids and exhibited a surface which conformed very closely to that of the mold.

As a result of the interest in fission products for use as heat sources and/or radiation sources, in early 1960. Hanford was authorized to supply a semi-refined cerium-144 rare earth product to Oak Ridge National Laboratories. Facilities were installed for collecting this product and a cask war designed for shipping this product as a stabilized, dry powder. Since that time, numerous improvements have been made in the design of the HAPO I cask as described in Revision O of this document. In addition, the development of strontium-90 as a major product requires increased flexibility in the utilization of these casks.

The Fermi age of thermal neutrons in a lattice containing both water and graphite in some sort of weighted average of the age in water and in graphite independently, with correction for volumes of non-moderating elements or voids. The correct weighting function has been in question during design calculations for the NPR. This paper presents a very simple and direct approach to the problem, resulting in a weighting equation which seems to be theoretically sound. Unfortunately, simple theories cannot be guaranteed to give good results in complicated systems; nevertheless, it appears that the weighting dunction derived here is to be preferred over methods involving empirical relationships which are of doubtful validity in the NPR geometry.

The purpose of this investigation is to make a preliminary study and evaluation of the application of ultrasonic equipment and associated handling equipment to the problem of decontaminating reactor tube dummy pieces. Reactor Section is currently planning the construction of batch process nitric acid facilities for dummy decontamination. It was felt that use of ultrasonics together with a solution less corrosive than nitric acid might provide adequate decontamination at lower project cost than the presently proposed nitric acid facilities. This report gives the progress of the investigation to date.

Measurements were taken to provide experimental evidence for determining the feasibility of a proposed method for converting Pu240 to 241 by exposing shielded plutonium to selectively transmitted pile neutrons.

The remodeling, fabrication, and instrumentation of a high vacuum system are described. The faults encountered with the old system and their solution are outlined. Tests were made on diffusion pump oils and on the reduction of pump oil back diffusion. The measurement of high vacuum pressures by a cold cathode ionization gage at various points on a vacuum system are discussed.

This report refutes earlier calculations made in the flux level of exponential piles. The results of the earlier calculations made on subcritical assemblies using critical assembly equations are in error. The calculations used in this report were more successful.

In view of the increasing number of methods for determining plutonium in terms of weight rather than radioactivity, a need has arisen for a plutonium compound that can be used as a gravimetric standard. An easily prepared compound that dissolves readily in nitric acid is desirable. The present study was made to determine the feasibility of using plutonium triflouride as a plutonium standard. The preparation, composition, stability, and ignition of the compound are discuss, as are certain properties of the plutonium dioxide resulting from ignition of the trifluoride.

As part of an investigation of solvent stability, a direct analysis of recycled Uranium Recovery Plant solvent* has been made to determine the type, source, and properties of the impurities which have been generated in the solvent and are present in the solvent after it has been in use for some time. This work has been directed towards the separation and identification of the impurities by compound class, and the conclusions are set forth in this report. Another purpose of this report is to present the separation scheme employed, which will likely find similar application.

Eight zirconium and eight Zircaloy-2 process tube samples were irradiated in the "H" test loop, Tube 0961-H, so that the corrosion tendencies of the material might be evaluated when exposed to radiation. The samples were delivered to the Radio-metallurgy Building in January, 1955, and examined as requested by C. D. Wilson of the Pile Development Unit, Pile Technology Section.

Nuclear safety specifications for storage, banding, or transfer of enriched uranium slugs depend upon slug arrangement and geometry, the presence of a moderator such as water or wood, and the presence of other fissionable materials and reflectors. Specifications are given for enriched I and E uranium slugs, solid uranium slugs of two types, and aluminum-uranium alloy slugs.

A corrosion testing program has been initiated in Purex and Redox storage tanks to obtain corrosion data on SAE 1020 carbon steel, two alloy-high strength steels, and an alloy steel in neutralized process wastes.

The discharge of cooling water from the Hanford reactors introduce radioactive contaminants to the Columbia River. These materials may subsequently bring about exposure to human populations either through the direct use of the water for sanitary purposes or transfers of the radioisotopes into the food chains. It is therefore desirable to keep to a minimum the amounts of radioisotopes released to the river.

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.

Pressurized water nuclear reactors may be designed based upon either of two concepts: (1) pressure vessel, wherein the entire core is placed in a large, high strength fuel channels within a low pressure container. The Plutonium Recycle Test Reactor is a pressure tube type reactor. Selection of this basic type of pressurized water reactor depended to an appreciable extent upon the availability of suitable pressure tubing.

A prototype instrument has been designed and built which demonstrates an ultrasonic pulse echo technique of measuring metal thickness in the range .010 to .080 inches. Extension of the range of measurement on the thick side appears straightforward, however, transducer frequency response considerations may limit the extension of the method for very thin sections. The instrument is based upon the principle that sections, when driven by pulses of vibratory energy, resonate at frequencies dependent upon the thickness of the section.

To forestall the oxidation of graphite in HAPO reactors, a process specification provides that the oxygen concentration in pile gas should be kept to a minimum. Close control in maintaining this limit is difficult when only periodic samples are taken because laboratory results are always obtained long after any required corrections should have been made. A continuous-type analyzer that provides immediate knowledge of the oxygen concentration eliminates these time delays.

A perspective drawing of typical hood cells arranged in series for an assembly line operation is presented. Advantages of this improved enclosure for low level radiation work are listed which recommend its use on new facilities in preference to more conventional designs.