From July through December, a total of 1750 hr of computer time was used by programmers in "debugging" and in running problems. With the acquisition of a second operator, the evening shift was initiated. A night-shift operator is presently being trained, and third-shifts operations will probably begin after completion of the magnetic-tape memory. / Engineering time is regularly scheduled for 4 hr each morning and 1/2 hr late in the afternoon. An electronic technician is on duty during evening-shift operations. / Installations of the magnetic-tape memory units is complete, and the units are expected to go into operation in the near future. Work is continuing on the new input-output system.

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 …

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.

The dose of ionizing radiation received by passengers and crew members traveling on commercial airlines which transport radioisotopes is of interest to all persons who travel the airways frequently. Tariff regulations permit the air transport of 40 units of radioactive materials where one unit represents a quantity of radioisotopes which when packaged delivers a dose rate of one mr/hr at a distance of one meter from the center of the package. Using five different types of commercial aircraft, shipments of radioisotopes were simulated and dose levels determined at various locations in the aircraft under conditions normal to commercial travel. The survey indicates that the tariff regulations should be amended giving consideration to the type of aircraft involved.

Portions of the NaOH-NaH phase diagram were studied by means of differential thermal analysis. Under certain conditions NaH will either react with NaOH according to the equation NaH + NaOH in equilibrium Na/sub 2/O + H/ sub 2/ or thermally dissociate according to the equation NaH in equilibrium Na + 1/2 H/sub 2/. Both of these reactions are suppressed by a high H/sub 2/ pressure; and NaH neither reacts nor dissociates to an extent sufficient to affect the results reported. This was evidenced by the fact that changing the H/ sub 2/ pressure above the system in the range indicated did not change, within the limit of error of the experiments ( approximately plus or minus 5 deg ), the temperature at which a phase change was started or completed. Therefore the concentrations of Na and Na/sub 2/O present in the samples must have been small in all cases. Under a high H/sub 2/ pressure, therefore, the system may be considered as essentially binary, consisting of NaOH and NaH. (auth)

Sodium hydride dissolves in and reacts with molten NaOH to give an equilibrium mixture of NaH, NaOH, Na/sub 2/O, Na, and H. In the case where there is a gaseous phase (hydrogen) and only one condensed phase, the system is defined by the temperature, pressure, and one composition variable. The equilibrium, H/ sub 2/ pressure, which is a measure of the H/sub 2/ activity within the melt, was determined as a function of the composition of the condensed phase(s) at 600, 700, and 500 deg for equilibrium mixtures with original compositions of 2.5 to 97.5, 5.0 to 95.0, l0.0 to 90.0, and 20.0 to 50.0 mole% NaH-NaOH. The equilibrium H/sub 2/ pressure-composition isotherms obtained by removing measured increments of H/sub 2/ were reproduced by reabsorbing H/sub 2/. Results for the 5.0 mole % NaH mixture were duplicated by starting with an equivalent quantity of either Na in NaOH or Na/sub 2/O in NaOH, and reacting with measured increments of H/sub 2/. The system is discussed in relation to the interdependent reactions involved, the phase rule, the thermodynamics of certain reactions, and experimental techniques employed. (auth)

A 4-ft-7-in. adsorption column packed with 1/8 in. sodium fluoride pellets was designed to reduce the hydrogen fluoride content of the fluorine being used by the Volatility Pilot Plant from 5% to less than 0.01%. It will be a non-isothermal packed bed with the bas inlet heated to 100 C to avoid plugging and the exit cooed to 25 C for more complete HF removal.

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.

Five uninsulated semipermanent freeze coils, series connected, can be frozen when submerged in 70 F water with a Freon-11 flow rate of 1.85 gpm at an inlet temperature of -40 F. The refrigeration unit of the HRT-CP is capable of delivering >3 gpm to a similar semipermanent freeze coil system located in Cell C. Therefore the number of F-11 risers required in Cell C of the HRT-CP can be minimized by series connecting this many semipermanent freeze coils where required.

Calculations are presented of the fast neutron dose rate, the gamma-ray dose rate, and the thermal neutron flux along the source plate axis in the Lid Tank with the tank filled entirely with water. These calculations are compared with experimental measurements. The calculations are absolute and are compared with absolute measurements.

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.

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.

A study was conducted to demonstrate the stripping and evaporation steps in a Purex-type uranium recovery process and to ascertain the operating behavior of the equipment under a range of conditions. The factors considered were control, effectiveness, and reliability of equipment and optimum feed point. Experimental procedures are described, and recommendations for equipment modifications are included. (J.R.D.)

The friction characteristics of a full-scale straight-tube model of the ART fuel-to-NaK heat exchanger were determined experimentally. The presence of the spacers resulted in a transition to semi-turbulent flow at a Reynolds modulus of 350. this semi-turbulent flow persisted up to a Reynolds modulus of 5,000. The circumferential spacers were found to contribute slightly more than the radial spacers to the pressure loss in the heat exchanger.

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.

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 Cr (VI) was autoclaved with rocking at 280 C for about eight days in the HB-5 facility of the LITR. The average corrosion rate established by the a 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 a correspondingly rapid corrosion.

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.

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.

A study has been made of the radioactive nuclides present in the cooling water of the Low Intensity Test Reactor. Prominent nuclides which have been identified include Np239 and certain fission products. The experimental evidence points to the probability that there could be adequate U235 contamination on the fuel plates to quantitatively account for the fission products identified in the LITR cooling water. inasmuch as new aluminum does not contain uranium, it must be assumed that the contamination occurs during the fabrication of the fuel elements.

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.

Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types, have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of rare gases, iodine, bromine, cesium, and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50% of the rare gases, 33% of the iodine, 9% of the cesium, and traces of strontium. After 25% burn-up, the cesium value increased to about 60%. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2% of the iodine, 10% of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15% burn-up, when melted at 1850 C, released up to 95% of the rare gases, 90% of …

A series of two-group calculations has been made on the Oracle for the purpose of obtaining critical-mass and flux distribution data for various ORR core configurations. The 3G3R code of Bate, Einstein, and Kinney was used, together with the RSP code developed by Nelson. This made it possible to obtain results for the three-dimensional case. The results, which are presented graphically, are intended to serve as a guide for the design of experiments until such time as actual measurements are available. The calculations were performed for the "clean cold" case, and it should be realized that the presence in the core of experiments and of fission products built up during operation will materially alter the flux patterns found. It is believed that the critical-mass data are accurate to within 10%. Within the fuel region it is believed that the thermal-flux patterns are the also accurate to this degree. Comparison of the results with MTR critical experiments, however, indicates that the thermal flux in the reflector in the vicinity of the fuel-reflector interface may have been underestimated by a factor of as much as 1.3. It should also be recalled that in a two-group calculation the "fast flux" is often a …

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.