This report analyzes Parian and Teflon that were exposed to solutions of 1AX, 1AS, and 1AF, ANL June 1, 1948 Flowsheet, under conditions described in the following report.

From introduction: "This report is concerned with exploratory experiments designed to studies the efficiency of a number of scavengers for zirconium and niobium adsorption and to provide preliminary information on the best conditions to employ on the most satisfactory of those scavengers studied, namely Super Filtrol FO. The data in this report include data obtained through October, 1949, and are by no means considered to be final. The investigation is currently being actively continued."

The following report provides results from laboratory tests on the corrosion resistance of various types of stellite facings on astenitic stainless steel T-347 to Redox streams.

The following report describes a chemical procedure to analyze fission products and long-lived rare earths in both untreated urine samples and the supernates from the precipitation steps in the TTA procedure for plutonium.

Introduction: "A limited test progress for the evaluation of type 347 stainless steel pipe which failed the Huey (?) test has been complicated. The tests reported herein were similar in scope to those reported in document number EW-20765, Corrosion Tests on Stainless Steel Plate, Waste Metal Recovery Process. Material tested was obtained on order number EWC-8553, National Tube Company Heat #3X9741. Lots 9789 and 9790."

Abstract: "A titrimetric method for the determination of fluoride in plutonium is reported. Prior to a steam distillation of the fluoride, the bulk of the plutonium is separated by precipitation of the sulfate. The fluoride in the distillate is titrated at a pH of 3.2 with 0.001 M thorium nitrate, using chrome azurol-S indicator. Applied to samples containing 1 to 15 parts per million fluoride, the recovery is 87% and the precision on the 95% confidence level is +/- 0.65 ppm."

The purpose of this report is to present the information obtained relative to the characteristic adsorption of radioisotopes in a long (forty feet) laboratory soil column treated with a high salt, process waste of high activity density. The movement of adsorbed cations in a soil column subjected to prolonged leeching by water also was determined.

Two methods for calculating the *reduction in η due to spectral hardening have been used. In the first method it is assumed that the absorption cross section of U-238 is l/v. This indicates no change in η within the error of the calculation. In the second it is assumed that the scattering cross section of uranium is constant. A decrease in η of 3.0% at the centerline of the slug is indicated by this calculation. The assumption U-238 is a l/v absorber is regarded as the most reasonable of the two.

The erection of a concrete biological shield at Hanford, Washington presented several unique construction problems. This shield encloses the radioactive core of a reactor and is designed to protect operating personnel from the harmful effects of pile neutrons and associated gamma radiation. Permanent steel forms were erected to close tolerance around the reactor and filled with high density concrete by means of prepacked method. Prefabricated sections of the steel forms, or crates, were five to eight feet deep and weighed up to 18 tons each. During erection, the process holes through successive crates were kept within 0.012 in. of their nominal distance. The left side, top shield, and right side of this biological shield were constructed integrally to form a rigid frame. The front and rear crates, which contain numerous process holes, were keyed into the side and top shields so as to permit relative expansions of shield components and to provide for their lateral support. These massive steel and concrete walls provide considerable resistance to blast and earthquake forces, support numerous process facilities, contain pile gas, as well as protect personnel from nuclear radiation.

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.

This report describes the evaluation of physical properties of AGHT and 185-W graphites that are important to pile construction and operation. On the basis of these data, purified AGHT graphite was allocated to either filler block positions in the central regions of the K piles or to the upper or lower reflectors. This decision was based on the similarities of physical properties between AGHT graphite and the other moderator components. A similar allocation could be made for 185-W graphite for some future pile if it can be successfully purified or if purification is not required.

A large (size 2) Purex pulse generator was installed in the 321 Building Tan Farm to pulse solutions in the prototype Purex HA Column. Flow sheet considerations indicated a need for information on leakage rates with various size weep holes in the pulse generator piston. In addition to leakage tests, experiments were conducted to determine the air required under the piston to hold the column contents away from the pulse generator piston thereby preventing leakage during shut down and the determine the time required to bleed the air trapped in the pulse leg when the column is first filled. [...] Leakage past the piston was not appreciably affected by pulse frequency. [..] Bleeding the air out of the pulse leg under start-up conditions, even with the piston weep hole plugged, was rapid. The longest time required to bleed the pulse leg was 36 minutes, which included a column-filling time of 30 minutes.

This report is a survey of methods for determining trace amounts of nitric oxide and nitrogen dioxide in air. The methods are evaluated in relation to their applicability to continuous monitoring procedures for these substances in stack gas and in air for human consumption.

The purpose of this paper is to examine the theory and results of chemical exchange experiments for separation of heavy metal isotopes. The aim is to find a means of predicting those chemical exchange systems likely to produce large isotopic separation factors.

The absorption of the important fission products by plants is quantitatively presented as a concentration factor which is defined as the ratio of the fission product concentration found in the leaves to the fission product concentration found in the nutrient substrate. Of the fission products, the isotopes of strontium were found to be the most important by virtue of their high concentration factor, long half-life, and low maximum permissible amounts for animals. Iodine and barium follow in importance, with cesium moderately important in some soils. All other fission products have concentration factors less than strontium by 100 or more. The effect on the concentration factor of different agricultural plants, as well as different organs of the same plant, causes variations of about a factor of ten or less for each isotope. The concentration factor tends to increase as the pH of the nutrient substrate is decreased. Addition of stable carrier to the substrate does not decrease the amount of the radioactive isotope that is absorbed into the plant. The presence of iodine and yttrium carrier actually causes a significant increase in the concentration of the respective radioactive isotopes in the plant tissue.

A study of the fabrication failures for zirconium and zircaloy-2 process tubes was made. In the tube reducing operation, a non-uniform reduction in area for the cross section was found to be a major cause of failure. In annealing studies, a cycle of 2 hours at 820 deg C in vacuum followed by furnace cooling produced the greatest ductility of extruded zircaloy-2 for the annealing treatments studied. The ductility of cold worked and annealed zircaloy-2 was found to be superior to that of extruded and annealed material. The strain rate of a cold working process was found to affect the ductility of zircaloy-2.

Metallographic examination of an irradiated powder metal uranium slug from PT-313-105-4M⁽¹⁾ was undertaken primarily as an evaluation of recently remotized metallographic equipment. It was, also, proposed to compare the grain size with values obtained ultrasonically on the same slug. It was further proposed to make a comparison, so far as limited data permitted, of post-irradiated powder metal uranium grain size with that to be expected from normal uranium similarly irradiated.

In this document, simple modes are used to demonstrate the importance of different material effects on predicted material behavior. This document should help one understand the basic assumptions made in stress analysis.

The effects of variables on the rate of Purex solvent deterioration were investigated with the emphasis on those deterioration products which cannot be removed by carbonate washing. The deterioration rate was found to be directly proportional to the acid and nitrate ion concentrations, and proportional to the square root of the concentration of nitrite ion. Other observations on the effects of temperature, relative merits of brand name solvents, and the effects of radiation are included. A method of calculating the equilibrium level of the solvent deterioration products based on study conclusions and various estimates is outlined.

For some time, it has been the opinion of the personnel of this laboratory and other investigators that an appreciable amount of the corrosion observed on stainless steel in nitric acid solutions might be due to the presence of some of the lower oxides of nitrogen; NO-, NO-2, etc. If this assumption is correct, the elimination of these compounds from process solutions should result in a significant increase in the service life of equipment handling nitric acid, such as acid fractionators. Since the corrosion problems experienced in acid fractionators and concentrators are very severe, an investigation of this theory becomes highly desirable. The purpose of this report is to summarize the work performed, to date, by this laboratory in the investigation of the inhibition of nitric acid corrosion of stainless steel.

With the construction of nuclear reactors, a great deal of emphasis has been placed upon studying the effects of radiation on the solid materials of construction. It has become necessary to consider radiation exposure as an environmental entity along with others so established such as temperature, electrical and mechanical stress, chemical resistance, weather, etc. Problems concerning the study of the behavior of materials that are thus increased when radiation exposure has to be evaluated singly or in combination with one or more of the other environments. This paper presents the general plan of attack being followed to solve such problems that arise when plastics and elastomers are placed in radiation fields. These studies were undertaken for two primary reasons: (1) to obtain engineering data for practical applications of the desirable properties of these materials for such devices as o-rings, gaskets, pipe, and electrical insulation; and (2) to use these data to enable a better understanding of the nature of radiation effects on these materials. The initial phase of the program is the screening of materials by subjecting them to various degrading environments and recording property changes. Thus, with data on how these materials performed, more comprehensive studies could be made …

When water is used as a coolant in any heat-producing process, the purity of the cooling water is of considerable importance, both from the standpoint of build-up of deposited solids inside the cooling tubes, and as an indication of corrosion of the tubes or any other materials with which the water comes in contact. The first problem has long been recognized, and is generally solved by pretreatment of the water. Efficient treatment can reduce the total solids content to less than 0.1 ppm, and the concentration of individual elements to the order of 0.01 ppm. If water of this purity is used, the analysis of the input and output stresses can result in some useful information. The input stream analysis, of course, is direct measure of the quality of the original cooling water, and frequent analysis by a reasonably fast method can be used to keep pretreatment under control. But of even greater significance is the difference in the impurity content of input and output streams. In a simple, straight-through system the difference generally will be negligible. If a closed, recirculating system is considered, however, with the coolant water circulating through the process to be cooled and then through a …

Twelve uranium-magnesium fuel material samples have been irradiated in the MTR at the request of the Pile metallurgy Unit. These samples were 0.40 inch in diameter by 1.5 inches long and were canned in Zircalloy-2 capsules. The uranium used in these specimens was in the form of chips which packs about 50 volume percent. Six of the samples contained a matrix of pure magnesium and the other six contained an alloy matrix of magnesium - 1.4 weight percent silicon. Two specimens of each matrix material were irradiated to 1000 MWD/T and a like number to 5000 MWD/T. Bend tests were performed on the samples and on unirradiated control samples to secure a measure of the effect of radiation exposure on the physical properties of the material.

Prevailing dimensional tolerance for sintered UO₂ reactor fuel components to be enclosed in metal jackets are expensively small. For instance, the PWR fuel pellet, as of November, 1955, consisted of a centerless ground UO₂ cylinder 0.3560" (+0.0005", -0.000") in diameter and 0.3527" (±0.0008") long. This pellet was designed to fit in a Zircaloy tube 0.3585" (+0.0005", -0.0000") inside diameter. The use of UO₂ shapes as pressed, or extruded, and sintered with diameter variations controlled to a tolerance of ±0.003" to ±0.005" would represent a fabrication cost saving worth considering. It is reasonable to assume that the annulus between a sintered UO₂ shape and its container will vary as a function of time of irradiation. Shrinkage from the can walls occurs with relatively low density (i.e., 85% of theoretical) sintered irradiation (cf. MTR Test GEH-4-3C). Contrary to this effect, the thermal expansion coefficient of UO₂ is about twice that of Zircaloy. Of possibly greater significance is the inevitable fractuce of dense UO₂ due to thermal stress. Such cracking effectively relocates the annular heat transfer gap to the interior of the oxide where the higher temperatures enhance the heat transfer coefficient. Thus, possibly the worst condition, with respect to heat transfer, is …