Abstract: "The sensitivity to fast neutrons of a CP Meter ionization chamber of the type used for health physics beta and gamma survey measurements has been investigated."

Introduction: "This report covers progress on the testing of the General Engineering and Consulting Laboratories Turbine Pump No. 1 from May 1, 1948 to Jun 15, 1948. All testing prior to April 30, 1948 has been described in Redox Experiment Testing Report No. 1 (Document HW-9474) and No. 2 (Document HW-9694)."

The following report studies an estimation of the amount of Pu(III) and Pu(IV) present in the IBP Stream, with the purpose of establishing an analytical, method suitable for routine control analysis. A method is discussed that is based on carrying Pu(IV) upon a zirconium phanylarsonate precipitate.

Abstract: "A method has been established for the estimation of volatile organic acids in aqueous process solutions containing UNH, nitric acid, ANN, sodium dichromate and small amounts of hexone. The practice is to distill a 400 ul or 500 ul sample in the presence of an excess of phosphoric acid and ferrous sulfate under a high vacuum; a special apparatus utilizing a receiver cooled with a dry ice-isopropanol mixture is employed. The distillate is taken up in isopropanol and then titrated potentiometrically with standard potassium hydroxide solution. Since nitric acid and organic acids are present, two end points are observed. The potassium hydroxide added between these end points is equivalent to the organic acids. In the titration, CO2 from the atmosphere or in the potassium hydroxide is a source of error. The former was avoided and correction was made for the latter."

Abstract: "A new method of determining 8-hydroxyquinoline by titration with electrolytucally-generated bromine has been developed. Results show good precision in the range 0.4 mg. to 1.6 mg. total 8-hydroxyquinoline, with 99% confidence limits of less than +/- 2%. An electronically-regulated constant current source, required by the method, is described. A new indicator system with high sensitivity to free bromine is discussed."

From statement of objectives: The purposes of the work described in this report are to provide data useful for designing and operating the dissolver solution clarification equipment to be installed in the Redox production plant.

The following document describes data collected from a 14-month period of analyzing crustaceans containing radioactivity. Provided is information based on the change of radioactivity through seasonal changes.

From summary: "The relative absorptions of X-rays by various substances are listed as a guide to the application of the X-ray photometric method in chemical analysis and to the estimation of interference by extraneous substances."

This report follows an investigation of a cargo truck found contaminated by various radioactive material on a routine trip from Arco, Idaho to the Hanford plant on May 8, 1952. The report analyzes the truck's route in an attempt to find the location of the contaminating region.

Introduction:"A corrosion test program in support of the Redox Project was carried out during the second quarter of 1950 to determine the corrosion rates of several austenitic stainless steels in simulated 202-B Waste Storage solution containing HF, HN03, and H2SO4 acids. The results of this program are contained herein and all original data are recorded in laboratory notebook HW-3282-T."

Abstract: An investigation was conducted to determine the absorption and translocation of fission products by Russian thistle from a localized spot of contaminated soil. The amount and identity of the radioactive elements absorbed and translocated by the Russian thistle is given along with the location of these elements in the plants. Beta radioactivity to the amounts of 10 microcuries per gram caused no visible effects on the growth habits of sectioned material are included.

The following report examines a rapid method used to eliminate the possibility of explosive perchlorate-organic mixtures in a control laboratory in order to determine radiocesium. These methods include the usage of chloroplatinic acid as a precipitate of radocesium.

The following report covers a test with the objective to demonstrate that the plutonium and uranium losses associated with the aluminum jacket dissolution could be reduced by substituting a water wash for the 5 per cent nitric acid wash following the coating removal and that this change would not adversely effect the product yield of decontamination in subsequent process steps.

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."

It has been suggested by many workers in the power reactor field that a more efficient utilization of natural uranium may be obtained in thermal reactors if the Pu produced in a fuel cycle were used to enrich a subsequent cycle in which the irradiation-depleted-uranium would be reirradiated. The work described here was done for the purpose of evaluating the reactivity value of the Pu enrichment under the assumption that all plutonium produced is, after chemical separation, fabricated into separate fuel elements and not alloyed with recycled uranium. It has been suggested that the reactivity value of the Pu decreases with exposure to such an extent that highly burned plutonium should be discarded. We, therefore, wish to look at two of the variables affecting the limiting exposure: the reactivity value of the fuel as a function of exposure and temperature and the fraction of potential fissions which would be discarded as a function of exposure. Though the residence time of the Pu fuel before reprocessing, refabrication, and recycling depends upon the relative decrease in specific power, decrease in reactivity due to fission product build-up, the lifetime of the fuel elements before failure, and the cost of reprocessing, only the first …

The Hanford Test Pile is a heterogeneous, low power, graphite moderated natural uranium pile. The pile consists of an eighteen foot cube of graphite containing 292 charged channels in a square array with an 8-1/2-inch lattice spacing. This report describes the reactor and the operating procedures used, and presents the latest results of calibrations performed. These calibrations supersede other work which has been done on the Test Pile and contain refinements of most of the early calibrations.

Past experience has shown that the demand for increase separations plant capacity comes up very regularly. One of the variables which greatly affects plant capacity is cross-sectional area of the individual vessels. Larger areas permit greater flow rates as well as more space for the installation of heat transfer piping (shell and tube concentrators). Design considerations of the separations plants vessels have been based on both circular cylinder and slab geometries. A study has been made to determine other vessel geometries that will result in safe vessels from a nuclear safety standpoint and at the same time offer larger cross-sectional areas than right circular cylinders. Vessels of elliptic as well as annular cross sections have been considered. It is neither the intent of this study to discuss the effects of intersection, vessel piping, etc., nor the pros and cons of fabricating feasibility and structural strength of these different shaped vessels. The main purpose is to make comparisons of cross-sectional areas (capacity parameter) of safe vessels so that vessel shape may be evaluated as one of the parameters in any design study for separation plants.

Approximately 15,000 "P" Process Aluminum caps clad with aluminum-silicon on the face and sides have been successfully canned. These caps were fabricated by the Aluminum Company of America from clad plate stock. An increase in canned slug yield of about two percent was realized from using these caps and no significant process problems indicated that the additional cost of clad plate might be off-set by the elimination of the centerless grinding now required on the side of the cap base. Therefore, processing of at least 100,000 additional aluminum-silicon clad caps is recommended, and consideration should be given at this time to completely converting from the present cap design to the aluminum-silicon clad cap.

On May 2, 1955, a request for comments on the construction and design of proposed continuous calciners was submitted to this unit by W. M. Harty. This request was complied with and a letter containing comments and suggestions on the design and construction of these calciners was written to him on May 6, 1955. The major suggestions were: 1) that the shell of the calciner be made as thin as possible, using external ribbing for structural strength, and 2) that the material of construction be type 304-L stainless steel welded with type 308-L rod. The recommendation regarding design was based on work performed by Collins, Krivobok, and this unit, which indicated that thick sections, on the order of one inch, tend to crack when subjected to thermal cycling. Type 304-L stainless steel was recommended as the material of construction because of its resistance to weld and base metal cracking, and because recent investigations indicate that a neucleation of precipitated carbides occurs after extended service in the sensitizing temperature range. On May 25, 1955, this unit received a letter from W. M. Harty stating that, on the basis of design considerations, the calciners would be fabricated from one inch thick type …

The determination of various parameters such as thermal utilization, resonance escape probability, control rod calibrations and etc. is often accomplished by the measurement of the pile period due to some given perturbation. With the present thoughts being directed towards plutonium fueled reactors, it was felt worthwhile to discuss a few of the basic differences in pile neutron behavior due to the smaller fraction of delayed neutrons per fission of Pu239 and the larger fraction of delayed neutrons per fast fission of U238 than occurs from fission of U235 alone.

Much attention has been given to the application of zirconium and zirconium alloys to process tubing for reactors. The properties of this material which make it attractive for such use are, as compared to aluminum: lower neutron absorption; higher design strength; and better corrosion resistance in hot water. Installation of process tubes in HAPO reactors requires that gas tight and water tight joints be maintained between the gunbarrel and process tube, and the process tube and nozzle, respectively. With aluminum tubes this is accomplished by forming a Van Stone flange at the ends of the tube and clamping the flanges between the gunbarrel and nozzle with suitable gasketing. No changes would be required in the existing pile components to install zirconium or zirconium alloy tubes if Van Stone flanges could be formed on these tubes. other methods have been tested, however, which eliminate the necessity for a flange. These will be discussed later.

The introduction of aluminum in hot process water has been studied extensively and numerous reports have been issued on the subject. The conclusions reached have been largely empirical and can only be used to state what is happening under the given conditions existing in the corrosion experiments. It is only with difficulty and a great deal of uncertainty that these empirical results can be used to predict what will occur under a different set of conditions. If the fundamental mechanism of aluminum corrosion in water were known, it is possible that more accurate predictions could be made and that steps might be taken to affect the controlling factors and thus reduce the severity of corrosion. To keep the system as simple as possible one would attempt to study the corrosion mechanism by corroding aluminum in pure water. However, this is undesirable because of the extremely low corrosion rate of aluminum in pure water and the large water supply and treatment facility required to supply sufficient high purity water for circulation about the corrosion samples.

Report describing the Hanford Atomic Works' Plutonium Recycle Test Reactor, its siting concerns, operation, and safety concerns.

Decontamination of stainless steel and carbon steel used in high temperature recirculation systems is currently being studied to obtain an effective and economical decontamination process for use in these systems. This report presents the preliminary investigation process which has demonstrated very effective decontamination and is low in cost.