The enthalpies and heat capacities of seventeen fluoride mixtures in the liquid state have been determined using Bunsen Ice Calorimeters and copper block calorimeters. The fluoride mixtures were composed of the fluorides of two or more of the following metals: lithium, sodium, potassium, beryllium, zirconium, and uranium. The enthalpies and heat capacities of most of these mixtures were studied in the solid state also. Estimates of the heat of fusion have been made. General empirical equations have been developed which represent the enthalpies and heat capacities of the fluoride mixtures in the liquid and in the solid state.

Changes made in the Thorex process flowsheet were a decrease in the extraction column acidity to decrease thorium losses and the addition of a second thorium solvent-extraction cycle to provide the increased decontamination required when thorium irradiated to 2000-4000 g of U233 per ton is processed. Bonded slugs could not be dissolved by the Thorex flowsheet procedure. Various laboratory scale studies on feed preparation, first-cycle variables, and radiation damage to the solvent are reported.

The countercurrent gas-liquid system BF3(g)—anisole·BF3(l) for the concentration of boron isotopes has been studied. The single-storage separation factor varies from 1.039 at 0°C to 1.029 at 30°C. Rate of exchange is rapid, and, with efficient contacting equipment, complete exchange may be obtained in less than 15 sec. A total separation of 1.525 has been realized in laboratory equipment. The critical-product reflux reaction is quite efficient. Only about 55 moles of BF3 remain in each million moles of effluent solvent under laboratory conditions. The vapor pressure of BF3 over the complex rises sharply as the temperature is increased. At 0°C the pressure is 150 mm Hg, and at 40°C the pressure has risen to 1800 mm Hg. From vapor-pressure measurements, an approximate upper limit of ΔH= -12kcal per mole of complex was calculated for the reaction [equation not transcribed]. Qualitative tests indicate good resistance of anisole to decomposition by BF3 under plant conditions. The uncatalyzed exchange of boron between BF3 and BCl3 was found to be too slow to be exploited in a countercurrent system. The single-stage, equilibrium separation factor for the Nitrox system is a function of acid concentration. At 26°C the factor ranges from 1.064 with 1 M acid …

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on circulating-fuel reactors and other ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided into three major parts: I. Reactor Theory, Component Development, and Construction, II. Materials Research, and III. Shielding Research. The ANP Project is comprised of about 530 technical and scientific personnel engaged in many phases of research directed forward the achievement of nuclear propulsion of aircraft. A considerable portion of this research is performed in support of the work of other organizations participating in the national ANP effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circulating-fuel type of reactor. The design, construction, and operation of the Aircraft Reactor Test (ART), with the cooperation of the Pratt & Whitney Aircraft Division, are the specific objectives of the project. The ART is to be a power plant system that will include a 60-Mv circulating-fuel reflector-moderated reactor and adequate means for heat disposal. Operation of the system will be for the purpose of determining the feasibility, and the problems associated with the design, construction, and operation, of a …

Observations of the effects of neutron damage to zirconium oxides led to the conclusion that the cubic form of ZrO2 is more stable to such damage than the monoclinic form. It has been reported that zirconium corrodes more rapidly in certain liquids when exposure is made under radiation (neutrons and fission products). It is well known that on heating monoclinic ZrO2 a transformation, monoclinic to tetragonal (very similar to cubic), occurs at about 1500°C. The transformation involves sufficient atomic rearrangement that pieces of ZrO2 normally crack and crumble. It is suggested that the effects of neutrons on monoclinic ZrO2 may be similar so that a protective oxide film on the metal would be destroyed soon after its formation. It might be possible, therefore, that the protective oxide film on zirconium metal which is normally monoclinic might be less resistant to corrosion under radiation damage than a similar film which was cubic.

Much of the shielding work carried out with the Bulk Shielding Reactor (BSR) has yielded data of particular interest for the design of swimming pool type reactors, However, it is often difficult for a reactor designer to locate such data since it may be recorded in a report primarily concerned with shielding problems. Therefore, this memorandum presents a bibliography of reports from the Bulk Shielding Facility arranged according to the application of data to the various aspects of reactor design.

Calculated data and graphs describing the effects of batch thermal-neutron irradiations on the buildup of fission products in natural uranium, enriched uranium, and thorium are presented together with empirical equations and plots correlating total fission product activities and/or decontamination factors. Fluxes of 1012-1015 are considered.

A fifth in-pile loop experiment, L-4-8, was completed. The loop operated in-pile for a total of 1637 hr, during which time the LITR energy output was 4377 Mwhr. The average fission power in the loop based o cesium analyses was 622 w when the LITR was at full power (3 Mw). Based on oxygen data, the generalized corrosion rate for the first 300 hr was 4.0 mpy; the rate for the remaining 1357 hr was 0.7 mpy. The nickel data gave parallel results. The corrosion of the type 347 stainless steel, Zircaloy-2, and Ti-55AX [unintelligible] exposed in the core and in in-line holders was generally consistent with that observed in previous in-pile loop experiments. Some differences with steel were attributed to the fact that this was the first loop containing steel specimens operated with 0.04 m H2SO4 present in the uranyl sulfate charge solution (0.17 m UO2SO4, 0.03 m CuSO4). Stress specimens, made from the alloys Zircaloy-2, type 17-4 PH stainless steel, and Ti-C-130-AM, were exposed in care, in-line, and pressurizer locations. Microscopic examination and average weight loss gave no indication of effects attributable to the stressed condition of the specimens.

Semiannual Progress report of the Oak Ridge National Laboratory Chemistry Division providing updates on various projects, experiments, and other work in inorganic and physical chemistry, nuclear chemistry, organic chemistry, chemical physics, chemistry of separation processes, radiation chemistry, and reactor chemistry.

Progress report of the Oak Ridge National Laboratory Biology Division providing updates on various projects, experiments, and other work. This report includes a summary of scholarly output from the division and departmental activities in cytology and genetics, mammalian genetics and development, microbial protection and recovery, mammalian recovery, pathology and physiology, general physiology, biochemistry, enzymology and photosynthesis, microbiology, plant biochemistry, and biophysics.

Progress report of the Oak Ridge National Laboratory Analytical Chemistry Division providing updates on various projects, experiments, and other work in ionic analyses, analytical instrumentation, radiochemical analyses, activation analyses, spectrochemical analyses, inorganic preparations, optical and electron microscopy.

Several potential hazards that have been recognized and anticipated in the design and fabrication of the pressure vessel in the Homogeneous Reactor Test are discussed. These hazards results from the high operating pressure and temperature of the reactor, the exposure of the reactor vessel material to potential embrittlement and other affects of fast-neutron irradiation, and the need for containment of corrosive flowing liquids. The steps taken in recognition of these hazards are also discussed. The applicability of present codes to the reactor vessel fabrication is considered. Additional fields are suggested where recommended practices developed by code writing bodies could assist in development-type reactor design and fabrication.

Progress report of the Oak Ridge National Laboratory Aircraft Nuclear Propulsion Project providing updates on various projects, experiments, and other work. This report includes summaries of project activities in: aircraft reactor test design, ART physics, ART instruments and controls, component development and testing, procurement and construction, ART, ETU, and in-pile loop operations, phase equilibrium studies, chemical reactions in molten salts, physical properties of molten materials, production of fuels, compatibility of materials at high temperatures, chemistry, analytical chemistry, metallurgy, dynamic corrosion studies, general corrosion studies, fabrication research, welding and brazing investigations, mechanical properties studies, ceramic research, nondestructive testing studies, heat transfer and physical properties, radiation damage, fuel recovery and reprocessing, critical experiments.

In order that fairly accurate thermal stress calculations can be made on the ART, it is necessary to have a reasonable picture of the temperature distribution in the reactor. To get the temperature distributions, and to determine cooling requirements in various parts of the reactor, one must know the heat deposition rates due to alpha particles, beta rays, gamma rays, and neutrons in all parts of the reactor. The present report contains only the basic physical data necessary to determine the heat deposition rates due to gamma rays. Neutron fluxes in the core and reflector regions of the ART are to be obtained from two-dimensional multigroup calculations (performed by the Curtiss-Wright Corporation). These fluxes, in conjunction with the neutron absorption cross sections, determine the neutron capture and inelastic scattering rates in the core and in the reflector. The data in this report permit the calculation of the number of gamma rays originating at various energies at every point in the core and reflector.

This report presents a summary of certain physical properties that have been determined experimentally on the fluoride mixture that have been formulated at ORNL (Rers. 1, 2). These properties include the density, enthalpy, heat capacity, heat of fusion, thermal conductivity, viscosity, Prandtl number, electrical conductivity and surface tension. In addition to the experimental data, values have been predicted for the heat capacity and density of the other mixtures from the correlations of these properties. Estimates of the viscosity have also been made for a number of the mixtures on which no experimental data were available.

Gram amounts of americium were separated quantitatively from kilogram quantities of lanthanum to yield an americium product approaching 90% purity. The remaining impurity was chiefly yttrium. Elution of americium from 25% loaded Dowex 50 resin column with 0.15 M citric acid— 0.10 M diammonium citrate — 0.3 M ammonium nitrate, pH 3.3 gave a product containing 99% of the americium with a La/Am ratio of 1/100 or less in one fourth of a column volume, in this case about 1 100-fold volume reduction. Approximately 9 g of americium was purified by this method. Elution with 12.8 M hydrochloric acid from a 20 to 30% loaded column gave 90% of the americium in two column volumes of product with a La/Am ratio of about 1/4. About 1 g of americium was purified by this method.

Calculated data and graphs describing the effects of continuous thermal-neutron irradiation of thorium, the usual method of operations of homogeneous reactors, are presented.The buildup and decay of U^233, Pa^233, other heavy isotopes, and fission products are considered on the basis of best available cross-section and fission-yield data. The effects of the heavy isotopes and fission products on neutron economy are discussed.

The enthalpies and heat capacities of seventeen fluoride mixtures in the liquid state have been determined using Bunsen Ice Calorimeters and copper block calorimeters. The fluoride mixtures were composed of the fluorides of two or more of the following metals: lithium, sodium, potassium, beryllium, zirconium, and uranium. The enthalpies and heat capacities of most of these mixtures were studied in the solid state also. Estimates of the heat of fusion have been made. General empirical equations have been developed which represent the enthalpies and heat capacities of the fluoride mixtures in the liquid and in the solid state.

Two revised Excer process flowsheets are presented for the production of UF4 from urnayl nitrate solution. In both, the uranium is sorbed on a cation-exchange resin and eluted with aqueous HF. In one, the resulting UO2F2 in dilute HF is mixed with dilute sulfuric or hydrochloric acid and electrolytically reduced in a cation-exchange-membrane cell, and the precipitated UF4·0.75H2O is dried and dehydrated to anhydrous UF4. A cost of 20¢ per pound of reduced uranium is estimated. In the other, ammonium fluoride is also added to the uranyl fluoride solution before the electrolytic reduction step, and the precipitate is UF4·04NH4F. A flowsheet for preparing UF4 from raw ore is also shown. The uranium is sorbed on an anion-exchange resin from a sulfuric acid leach liquor and eluted with dilute HCl. Ammonium fluoride and HF are added to the UO2Cl2-HCl solution before the uranyl ion is electrolytically reduced to precipitate UF4·0.4NH4F. Decontamination factors as high as 45 and 9 for gross beta and gamma activities and of 10 to 5000 for light metal impurities were obtained in the electrolytic reduction step.

Consideration has been given to various forms of fissionable material for use in atomic reactors, including the pure metals, their alloys and compounds. Of particular interest is the dioxide of uranium which is refractory and corrosion resistant in some environments.The oxide is useful in both granular and bulk forms. Small grains of uranium oxide can be mixed with other materials to form matrix type elements where they serve either as a convenient or necessary form of fuel or fertile material. For other applications the oxide may be fabricated in bulk form such as pellets, rods, plates, or blocks.There is a need for knowledge of the properties of the properties of this oxide, particularly as it affects fabrication in the various forms required. This knowledge is also required by reactor designers and engineers. There is in addition a challenging field for basic studies of sintering rates, oxidation behavior and other phenomena. Fabrication techniques have been developed to produce uranium oxide in various forms with consideration given to the economy of production. The continued application of basic knowledge of these materials has led to simpler. more practical means of fabrication and has thus widened the scope of their use in atomic reactors.

The methods of analysis given in this report are those which were used in the Analytical Chemistry Division of the Oak Ridge National Laboratory for analyzing samples which were derived from the experimental work on the separation of the isotopes of boron by chemical exchange. The samples consisted principally of boron trifluoride solutions in anisole (methyl phenyl ether, CH30C6H5). The boron concentration ranged from a few parts per million to 5 or 6 per cent. Boron was determined on all samples. During the early stages of the project, iron and copper were occasionally determined, while a limited number of aqueous solutions and water extracts of anisole solutions of BF3 were analyzed for fluoboric and hydroxyfluoboric acids, boric acid, total boron, and total fluoride. Boron was determined by the use of either a spectrophotometric or volumetric method, depending on the amount available. Initially, if the amount of sample and boron concentration were such as to provide a total of at least 2 to 4 mg of boron, the volumetric method was utilized and found to be satisfactory. For smaller amount, the spectrophotometric method was used. Later, because of its greater speed and simplicity, the spectrophotometric method was used for samples in …

In some recent experiments the ion density in a plasma was measured as a function of radial distance from a d-c arc source of ions for the magnetic field intensities from 2500 to 14000 oersted. The diffusion coefficient appeared to very inversely as the square of the magnetic field strength, D~1/H2. The absolute value of D was shown to be approximately that which would be predicted by the collision diffusion theory when account is taken of the shorting effect of the end walls. The purpose of this report is to continue the examination of ion diffusion in a plasma and field experimentally the relation between the diffusion coefficient and the neutral gas pressure for a constant magnetic field.

Electrodialysis in an ion-exchange membrane cell was shown to be technically feasible for the concentration of an azeotropic mixture of HF and water. A flowsheet is presented for recovery of anhydrous HF by distillation and electrodialysis of the azeotropic residue. In the electrodialysis step, 2.6 kwhr of energy per pound of anhydrous product was consumed, with electricity at 1¢ per kilowatt-hour, the total operating cost of the electrodialysis equipment alone, including amortization, would be 6¢ per pound of HF.

Thorium, a fertile material, is of interest to the Reactor Program in the production of U233. Thorium can be extracted and processed to a very pure bulk metal for fabrication into solid fertile elements. There are advantages, technical and economic, for using fabricated bulk thorium dioxide rather than the metal in some applications. It is the purpose of this paper to point out these advantages and to present briefly the technology related to fabrication, radiation damage and chemical processing of ThO2.