Report discussing experiments and developments in purification of graphite starting from March, 1949.

Report containing results of a study regarding concentration of radioiodine in edible plants fed to animals.

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."

Introduction: "A corrosion test program, to determine the suitability of LaBour B-55 and Worthite for use in the construction of circulating pumps for th UO3 process has been completed and the data are appended."

One of the UO2 fuel element designs proposed for use in the PTRT is the nested tubular concepts. This configuration compromises a central fuel todo surrounded by two concentric tubes of fuel (see sketch, appenx I.) . These UO2 shapes are to be jacketed in the Zircaloy and must be separated from each other and the procuresses tyvm vt annular spaces for the passage of coolant. The annuli are established and and maintained by the longitudinal ribs on the outer surface of all three jacketed fuel elements.

Two xenon extraction runs were made this month. It appears that a small design change in Trap #2 will be necessary so that a dry ice-trichloroethylene slurry can be used for coolant rather than liquid freon. For each of the runs this month the enriched generator was exposed for four hours in the est pile operating at 100 watts. A period of eight hours for cooling and xenon builidup was allowed before the collection and separation runs were started.

In obtaining the results reported here a double focusing magnetic type spectrometer was used. Momentum spread was about 1%, and in most cases 2560 counts per point were taken. An end window Geiger tube was used as a detector. The window was of mica with areal density of 1.6 mg/cm2.

Various organizations in the Biophysics Section of the Radiological Sciences Department participated in evaluating local radioactive particle fallout as influenced by the nuclear explosions at the Nevada Proving Grounds during the early part of 1953. Data collected from approximately 50 air monitoring stations operated by the Regional Survey forces were used to determine the trend of radioactive particle concentrations in the atmosphere between March and June. These measurements were supplemented with the findings of associated organizations in the Biophysics Section during and immediately after the peak influx on May 26.

The investigation to be described in this paper was one of several undertaken to resolve a curious discrepancy that had arisen in measurements of the diffusion length of graphite. Those diffusion lengths measured with the graphite layups of 100 Area piles were three to six centimeters higher than the value obtained with the Hanford Standard Pile. The fact that the 100 Area structures of the order of 40 ft. X 40 ft. X 30 ft, not including reflector, approximated an infinite medium, whereas sigma piles are of the order of 8 ft. X 8ft. X 10 ft., suggested the possibility that the diffusion length obtained by mensuration of relative fluxes and application of the conventional equations depended in some way on the size of the medium. Therefore, in order to determine if such a dependency actually exists, it was proposed to erect a series of graphite stacks of different dimensions and to make diffusion length measurements in each.

The purpose of this manual is to present available data which can be used to determine disintegration rates of radioactive samples from counting rates measured under given circumstances with routinely used laboratory instruments. There is also brief discussion on the methods by which the various factors have been determined with reference to the accuracy with which they are known and to what extent the finally quoted disintegration rate is affected by these errors. Emphasis has been placed on those isotopes of particular interest to Radiological Sciences programs and the manual is intended as a standard through the use of which comparable values can be reported from all units in the Department.

Uranium dioxide, uranium monocarbide, and uranium mononitride are potentially useful ceramic nuclear fuel materials. This paper reports the results of a determination of the melting points of these materials.

At the request of Design Development Operation, various heat transfer and fluid flow problems were examined which are peculiar to the CO2 Gas-Cooled Loop in the PRTR. The results of these calculations are desired primarily to aid in demonstrating the adequacy of the design proposal. In addition, the operational limits of the loop and the consequences of the gas loop installation on the PRT reactor are of interest.

One of the research studies of the Chemical Effluents Technology Operation is the improvement of the method for predicting the capacity of a crib for the retention of wastes. In addition to laboratory work the research was extended a field experiment using a simulated crib fed with a solution containing a radioactive tracer. The purpose is twofold: (1) to check the validity of laboratory findings and (2) to observe several features of the behavior of solutions put to ground in the field.

In order to prevent dangerous concentrations of combustible gases such as hydrogen and ammonia in Redox dissolvers, a continuous analysis of the off-gases is needed. Hydrogen is evolved from most fuel element dissolution processes now used or planned for use at Redox. Ammonia is also released from ammonium fluoride dissolution of zirconium and caustic dissolution of aluminum cladding. Oxides of nitrogen and hydrogen are formed during nitric acid dissolution of uranium or plutonium-aluminum alloys. Since no fixed ratio exists between the concentrations of any two of these gases, the determination of combustible gases in complex mixtures is not possible with a single bulk property analyzer.

The multi-purpose dissolver was designed to provide: 1. A critically safe vessel for dissolving uranium metal of enrichments up to one percent U-235. 2. A vessel which can be charged with fuel elements up to 10 feet long. 3. An off-gas arrangement which will eliminate return of ammonia to the dissolver during the Zirflex decladding process. 4. A vessel in which small charges can be dissolved without using extra chemicals. 5. A by-pass routing around the silver reactor during cladding removal.

It is well known that neutron damage generally causes increases in the yield and ultimate strength and a decrease in ductility of a metal. There is a continuing program at HAPO to determine the extent of these changes in Zircaloy-2 as functions of integrated neutron exposure, irradiation temperature, and reactor atmosphere. Three investigations from this program will be described and the results summarized. The first investigation deals with both annealed and cold worked Zircaloy-2 irradiated at approximately 50 C. and the other two investigations deal with annealed Zircaloy-2 irradiated at approximately 100 and 280 C respectively. In each investigation tensile testing was performed at room temperature.

The development of a miniature probe was desired for measuring approximate single nuclide beta dose rate in solution and in various animal organs. This probe designed for biological experiments, was to have maximum possible sensitivity to detect low levels of nuclide concentrations. The desired dimensions of the light pipe were to be approximately one-fourth-inch diameter with lengths of three to twelve inches.