Introduction: "The carrier (pyroelectric) concentration method is commonly used in the spectrographic analysis of plutonium metal. This report describes work which lead to substantial improvements in accuracy, precision and ease with which the analysis is made. Included are data from analyses of plutonium metal by the original and the improved carrier methods, and the cupferron extraction method."

Purpose: "This test was conducted to gain information and experience in the handling of heavy aggregate concrete. The results from this test were to establish procedures for a subsequent program -- "Construction Test of High Density Concrete Shielding.""

This review covers: (a) The continuous countercurrent ion exchangers which have been described in the patent and/or technical literature. (b) Some of the other moving bed processes which are similar in mechanical arrangement to a continuous countercurrent ion exchange and which might have certain features adaptable to an ion exchange unit. (c) The auxiliary mechanical devices which are or may be used in an ion exchange apparatus.

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.

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 …

A flexible foil data processing program is described. Raw data on foil radioactivity are the basic input information required. Output may consists of relative activities, saturated activities, and/or cadmium ratio and flux spectrum data. A statistical analysis of the data is executed with the direct calculation, and errors estimated for the output data.

Results of eight runs using in-line U photometers in organic and aqueous streams of the Purex Hot Semi-Works are presented. Their operation, both mechanically and electrically, was satisfactory, indicating changes in stream U concentrations over wide ranges.

Flowsheet considerations and information developed for the large (size 2) Purex pulse generator⁽¹⁾ indicated the need for information on the operating characteristics of the small (size 1) Purex pulse generator.

The purpose of this report is to describe the tests performed on the D-12 waste evaporator, to present and evaluate the data obtained during the test, and to make recommendations for the implementation and operation of present and future installations of bayonet tube bundles.

It is the purpose of this report to: 1. Present the results of a study to estimate the range of attainable boiling conditions in the 1706 KER loop design on the bases given in reference (12). 2. Present an outline of the significant modifications and equipment changes necessary to obtain the process conditions desired.

The effect of prolonged heating of cold worked Ziracloy-2 at a temperature of 360°C on the mechanical properties of the material was investigated.

An experiment now in progress should give some accurate information about the thermal neutron absorption cross section of carbon 12. This report outlines and summarizes this experiment and analyzes it to determine the main sources of error and the probably error in the final result.

Radiobiological studies were made to determine effects of radioactive effluents from the Hanford reactors upon the aquatic biota of the Columbia River and to evaluate related hazards. Data from studies completed between September, 1945, and December, 1955 are presented and interpreted. All forms of life were many times more radioactive than the water they inhabited. Some radioisotopes were much more readily accumulated than others in living organisms. Differences in the concentration of certain radioisotopes by various species of organisms and kinds of body tissue are described; and geographical, seasonal and annual fluctuations in the concentration of radioisotopes in organisms are discussed.

In discussion with a number of people at HAPO, KAPL and ANL a hypothesis has been evolved which appears to fit available information concerning irradiation embrittlement of uranium as well as indicate a possible solution to the problem. The purpose of this memorandum is to expound the hypothesis as an aid to those working with the problem. Since it imbodies the ideas of many people, no claim to unique authorship is implied.

The moderator fluid will be injected into the PRTR calandrin through injectors located between the shroud tubes and at the bottom of the calandrin. It is important that the size and arrangement of the injectors be such that complete mixing of the moderator will occur and prevent hot sports from forming in the moderator. Such hot spots could lead to undesired changes in the moderating characteristics due to boiling within the moderator. Also of importance is the requirement that the injector should not produce excessive turbulence at the moderator surface thereby complicating moderator level control. To determine the extent of moderator mixing within the calandrin, experimental studies were made employing a full scale PRTR calandrin mockup.

The small amount of irradiation tenting experience on the plutonium-aluminum type elements planned for the Plutonium Recycle Test Reactor (PRTR) has made such testing of great importance. The high temperature pressurized recirculating water 1706 KER facility is one possible place for conducting investigations of the irradiation behavior of this type fuel element. To obtain the maximum information from the in-reactor testing and to detect possible problems, out-of-reactor test both at room and anticipated operating temperatures must be made. Room temperature pressure drop measurements and high temperatures performance of two prototypical fuel element designs proposed for KER testing are reported in this document.

Several tests are being performed and others being planned to investigate the role of heat transfer in corrosion processes. These tests are measuring both corrosion rates of metals (Zr-2 and X-8001 aluminum) under heat transfer, and the temperature rise associated with the buildup of the corrosion product. A brief description of these tests is given in this report.

The wrist badge provides a dosimeter that is useful in estimating the radiation dose to the hands and forearms. Its new shield system gives good gamma and slow neutron dose discrimination with duPont 552 film packets. The film can be evaluated using the present technique and equipment. Several attempts to develop hand dosimeters have been made. Finger rings using film have been used routinely but have not been entirely satisfactory for all situations. The wrist badge was developed to provide improved gamma and slow neutron dose measurement of the upper extremities under certain appropriate conditions. The wrist badge dosimeter is not a substitute or alternate for finger ring dosimeters but is a necessary dosimeter for some extremity exposure situations.

A disposal method is needed for wastes generated from the decontamination of the NPR primary coolant loop. The limitations imposed by facilities design criteria for the disposal of NPR wastes preclude direct river release of the spent cleaning solutions because of the anticipated quantities of radioactive material in these wastes. The soil at a 100-N Area trench or crib should not be relied on for removing radionucleotides by ion exchange or filtration because of the high salt content of the wastes and the presence of solubilizing reagents. Permanent or long term storage of large volume of decontamination wastes would be expensive. A waste treatment is sought for concentrating the radioactive materials to volume suitable for long term storage and which would permit dispersal of the excess liquid to the environs.

Continued operation of F reactor with high water collection rates during the past 12 years has resulted in numerous detrimental effects. In addition to promoting external corrosion tube leaks, water leaks have corroded the Gunbarrel to the biological shield donut assemblies and cast iron thermal shield blocks, thus preventing the majority of tubes in F reactor from unrestrained thermal expansion. Fatigue of the Van Stone flange under the resulting compression loads leads to eventual failure in some cases. In addition, excessive compression loads exerted against the nozzle gasket result in plastic deformation and eventual failures of the gasket.

Storage of the fission products separated from the product streams of the Purex process is being accomplished using a smaller volume of accompanying solution than any other process here-to-fore used at HAPO. The operating technique and control mechanisms which are needed to store large quantities of these highly radioactive wastes are not yet fully understood, but considerable insight into the problem has been gained from the experience at Redox during the last 36 months. The basic intentions of the 241-A Storage Facility design is to control the boiling wastes by providing suitable tanks to contain the liquid and a vapor system provided with suitable seals to control the vapors. This document (Part I) will present a somewhat detailed description of the Purex Storage Facility and a review of the activities there before plant start-up. Part II, published under separate cover, contains a description of Waste Farm Technology including a process description and a recommended plan for operation.

Considerable interest has been shown at HAPO in the development of a continuous ion exchange process for concentrating plutonium solutions. Development work has been performed on continuous ion exchange for both uranium and plutonium concentrations at the X-10 at ORNL. On a recent trip to Oak Ridge to discuss critical mass problems and experiments with Dr. A. D. Callihan of the ORNL critical mass facility, a meeting was also held with C. W. Hancher and R. Higgins of X-10 regarding continuous ion exchange operation. From this meeting, information was obtained that is helpful to work out nuclear safety aspects of such a plant for the concentration of plutonium solutions. An advance copy of "Countercurrent Ion Exchange" by T. A. Arehart, J. C. Bresee, C. W. Hancher, and S. H. Jury was obtained. This paper is to be presented at an AIChE meeting this fall. Preliminary blue prints of ORNL-3" Ion Exchange column design were also obtained. Upon my return to HAPO, a meeting was held with members of the Process Planning, Equipment Development, and Chemical Engineering Development Units of the Chemical Engineering Sub-Section (Separations Technology) to present the document and blue prints on continuous ion exchange and arrive at 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.

Under the general heading of Reactor Lattices are reported a new method of calculating resonance escape probabilities, results for the effective mass of a proton bound in the water molecule, and a comparison of similar techniques for calculating lattice constants developed independently by Hanford physicists and by a school of Russian physicists (as reported at the Geneva Conference). An estimate of maximum errors in f and η for thermal systems is described. Experiments and analysis thereof on the effects of neutron streaming in air channels through a moderator are reported. Buckling calculations and experimental results are given for graphite lattices employing 1.17-inch diameter natural uranium slugs and U²³⁵-Aluminum alloy slugs. An experimental measurement of the critical mass of an annular cylindrical array is described. Some measurements on, and the status of, construction of the Lattice Testing Reactor are reported. Under Instrumentation, development of a BF/sub 3/ counter suitable for operation at elevated temperatures is described. Some measurements on, and the status of, construction of the Lattice Testing Reactor are reported. Under Instrumentation, development of a BF₃ counter suitable for operation at elevated temperatures is described. A method of determining the screening parameter in the Thomas-Fermi model of the atom is …