This quarterly progress report details the ongoing research and experiments at the Oak Ridge National Laboratory as part of the Aircraft Nuclear Propulsion Project. The first part of this report discusses reactor theory and design. The second part of this report is not included. The third part of this report discusses materials research. The fourth part of this report includes appendixes

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 …

Technical report covering experiments happening on the Analytical Chemistry Division's sites at the Oak Ridge National Laboratory. Includes information on ionic analyses, radiochemical analyses, spectrochemical analyses, service analyses, inorganic preparations, analytical chemical control of homogeneous reactor solution, optical and electron microscopy, and service analyses. [From Abstract]

In an elastic collision the neutron loses part of its kinetic energy to the nucleus with both the kinetic and momentum of the system being conserved. However, for many elements the scattering is not isotropic in the center-of-mass system at the higher neutron energies. Many of the present reactor multigroup codes include anisotropic scattering at the high neutron energies, while many others assume isotropic scattering at all energies. In order to consider some of the effects of including anisotropic scatting, reference is made t the multigroup equation generally assumed for the slowing-down density.

The information concerning beryllium assembled herein has been selected primarily to provide information that might be pertinent to the design of the ART. The creep-rupture data in the literature have been obtained from vacuum cast and extruded beryllium, rather than from the hot-pressed beryllium which will be used in the ART. A research program is being conducted by the Brush Beryllium Company, under contract to the Oak Ridge National Laboratory, which will eventually provide the design engineers with data on the high temperature strength properties of hot-pressed beryllium. In the interim, it will be necessary for the metallurgists and design engineers to make judicious estimates and extrapolations from the data available.

Studies are being made on the recovery of thorium (and uranium) from granitic rock, since this source represents a very large potential thorium reserve for the nuclear power industry. In preliminary leaching studies on 16 granite samples (containing 8-95 ppm thorium and 1.5-16 ppm uranium), maximum recoveries of thorium and uranium ranged 30-85% and 15-65% respectively, and sulfuric acid consumption was high (30-120 lbs H2SO4 per ton of granite). A relatively high acidity was needed to obtain rapid and efficient dissolution of the soluble thorium fraction. The cost of treating granite was estimated at $3.50-5.20 per ton, variations within this range being dependent primarily on differences in acid consumption for different granites. Estimated costs per pound of thorium plus uranium recovered ranged $30-500.

Technical report presenting a summary of the flowsheets, equipment, and progress for the Oak Ridge National Laboratory (ORNL) pilot plant development of the Purex Process and of the ORNL Metal Recovery Program. As of this report, conversion of the Purex pilot plant equipment in Buildings 3019 and 3503 is 90% complete. The building structure for the ORNL Waste Metal Recovery plant had been completed, and 15% of the process equipment had been installed. The first hot runs for this plant were scheduled for June, 1951. [From Abstract; Introduction]

Technical report covering chemistry of source, fissionable, and structural elements, nuclear chemistry, radio-organic chemistry, chemistry of separations processes, chemical physics, radiation chemistry, and instrumentation at the Oak Ridge National Laboratory for the period ending September 30, 1950. [From Abstract]

Factors affecting the results of thoria particle size distribution measurements by sedimentation procedures currently and recently employed are considered. The effects of thoria concentration, solvent, dispersant, thoria properties, and other factors are discussed.

Several compatibility test of various materials in contact in sodium under 500 psi pressure were conducted for 716 hr at 1500 F. Of the diffusion couples studies, the Inconel-beryllium system showed the largest amount al alloying. The reaction between molybdenum and beryllium resulted in the formation of two compounds, both of which were severely cracked in several areas. The molybdenum-INOR #8, and the INOR #8p type 316 stainless steel interfaces showed little if any alloying.

From abstract: "This report summarizes the corrosion data obtained for nine construction materials located throughout the ORNL Purex Pilot Plant Acid Recovery Unit during seven hundred hours of operation."

Report documenting the corrosion of fluorinators of the Oak Ridge National Laboratory's volatility pilot plant. Includes descriptions of the fluorinators, their operating environments, and the results of their corrosion.

Technical report outlining how uranium hexaflouride was prepared be the direct combination of irradiated uranium metal with elemental fluorine and subsequently decontaminated by adsorption, filtration, and sublimation on a laboratory scale. Report proposes a survey of other methods of preparing UF6 from uranium metal, a study of adsorption techniques for removing plutonium from UF6, and an investigation of fractional distillation for removing the volatile fission product fluorides from UF6. [From Abstract, Summary]

Technical report detailing corrosion tests of Type 347 stainless steel for two weeks in boiling uranyl nitrate, sulfuric acid media, with and without additions of Al+3 and F-1 in a 2:1 mole ratio. The results of these tests show that although the presence of aluminum greatly reduces the deleterious effect of fluoride, the corrosion rates are still three to four times greater than when no fluoride is present. [From Abstract]

It has been reported in the literature that prolonged soaking of 16 wt & uranium-aluminum alloys at elevated temperatures induces a microstructural changed termed "conglobulation". this process is essentially a spheroidization of the UAl4 eutectic. Conclusions of the tests: (1) Extended soaking of 48 wt $ uranium-aluminum alloy billets at 600 C does not improve their hot-rolling characteristics. (2) Soaking periods of 50 hours and greater reduce the tensile and yield strengths of these alloys. (3) Soaking for 50 hours results in complete spheroidization of the eutectic in 48 wt % uranium-aluminum alloys.

From abstract: "On the 86-inch cyclotron a 41-kw beam has been calorimetered at a net ion loading efficiency of 40%; the average proton current was 1.85 ma at 22.5 Mev. Practical specific yields have been determined for (p,2n) reactions on zinc and bismuth. The investigation of products of proton-induced fission of uranium has been continued and new techniques are being used in measuring angular distribution of reaction products. The 63-inch heavy particle cyclotron is now in operation; N+++ particles have been accelerated to ~25 Mev. In preliminary tests, induced activities have been detected in carbon, nitrogen, and oxygen targets. A hot-cathode ion source is ready for test operation. The 22-inch cyclotron is being used in an investigation of the problems associated with the use of rf and dc electrodes for the acceleration of protons from the ion source into the dees. Radiation-induced corrosion in Inconel tubing containing #21 eutectic (ANP) has been produced by proton irradiation in the 86-inch cylclotron; it is shown that the corrosion was not due to thermal effects. Approximately 150 grams of highly purified U 238 (< 5 ppm U 235) have been prepared, and two grams of thorium 230 (ionium) has been enriched to 90.6%."

Report that investigates the concept of using ammonium fluoride-oxidant systems to dissolve spent U-Zr-Sn reactor fuels in both batch and continuous operation.

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

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.

Report issued by the Oak Ridge National Laboratory discussing the development of a high-density concrete. As stated in the abstract, "the basic approach to high-density concrete mix design is reviewed to show that theoretical calculations for concrete mixes do give results that are reliable" (p. 1). This report includes tables, illustrations, and photographs.

Current status of the High Flux Isotope Reactor which is being planned for construction at Oak Ridge.

The essential physical and mechanical properties of materials which enter into the design considerations of high-temperature systems have been outlined, and the relationship of these factors to the detailed design of structural members noted. Particular attention is given to the phenomenon of strain-cycling and relaxation, and several engineering devices wherein these effects can markedly influence design are mentioned. Techniques and apparatus for the measurement of strain-cycling and relaxation properties are described and some recent data for Inconel at 1500 F is reported. A brief discussion is presented on design-criterial for high-temperature applications, and it is pointed out that in some high performance devices the considerations of the strain-cycling phenomenon may impose the design philosophy of "steady-state" operation. Some elementary applications of strain cycling data to design are described, and test experience indicates that analytical techniques can be developed for predicting with reasonable reliability of the fatigue life of structures.

The in-plie slurry loop work is now being considered as a joint program between the PAR project and ORNL. It is proposed that PAR design, fabricate and test the in-pile loops and that ORNL operate the loops in-pile, dismantle the loops after irradiation and made the appropriate measurements to determine the radiation effects. This report gives the objects of the slurry in-pile program and outlines the facilities and operations required to execute ORNL's part of this program.

In order to test the effect of chromate ion on the corrosion of Zircaloy-2, a 0.04 m uranyl sulfate solution (enriched) containing added acid, copper sulfate and 557 ppm of Cr (VI) was autoclaved with rocking at 280 degrees C for about eight days in the HB-5 facility of the LITR. The average corrosion rate established by the rate of consumption of oxygen was 9.9 mpy at a power density of 4.9 w/ml. It is indicated by comparison with a previous corrosion study under LITR radiation that the presence of Cr (VI) had no significant effect on the radiation corrosion of Zircaloy-2 by enriched uranyl sulfate solutions. However, the data are not conclusive and may be interpreted as showing a low corrosion rate for a limited period (i.e. a short term inhibition) followed by correspondingly rapid corrosion.