This quarterly progress report discusses ongoing research and experiments at Oak Ridge National Laboratory in the Metallurgy Division. This report discusses thorium research, the mechanical properties of metals including thorium, uranium, inconel, and z-nickel, the ANP program (including the following topics: the physical chemistry of liquid metals, and welding research), and the materials testing reactor program.

Tests were conducted to determine methods of removing scale deposits from the HRT heat exchangers. A mockup of the heat exchanger header was cleared of a deposit of iron rust by reverse flushing at a flow rate below 75 gpm. A tube bundle consisting of 109 1/4in. O.D. x 0.049 in. wall tubes was plugged with rust. Approximately 80% of these tubes were unplugged by using a 70 psi water pressure differential in combination with vibration from a pneumatic rivet gun. No mechanical method was employed in the tests which could clear the remaining tubes.

In connection with the program of the measurement of eta for U233, several critical experiments have been performed by R. Gwin and D. W. Magnuson of ORML with light water solutions of uranyl nitrate (highly enriched in either U233 or U35) in an essentially bare sphere 27 inches in diameter. This report presents the results of two multigroup-diffusion-theory calculations for the above experiments performed by C. B. Mills and associated at Los Alamos. Assumer cross sections, material concentrations detailed neutron balances and a comparison with elementary theory are included. The agreement between the calculated and experimental multiplication constants is excellent for the multigroup calculation but only fair for the elementary calculation. The latter method overestimates the fast leakage so that the computed multiplication constant is less than that found experimentally.

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.

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.

Design philosophy, status of component development, operational procedures, and limitations of each section of the Oak Ridge large-scale, high-temperature, thoria slurry facility.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Project. Progress of operations, design, development, and analysis is presented. This report includes tables, illustrations, and photographs.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Molten-Salt Reactor Program. Descriptions of the progress and studies conducted is presented. This report includes tables, illustrations, and photographs.

Report issued by Oak Ridge National Laboratory that discusses progress and work performed by the Chemical Technology Division.

Report documenting the ongoing research and developments of the Chemical Technology Division of the Oak Ridge National Laboratory.

Report containing the description and design of U₃O₈-Aluminum Dispersion Fuel Elements for Core I of the Puerto Rico Research Reactor. Topics include the requirements of core materials, materials specifications, manufacturing procedures, and shipping preparation.

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.

A partial bibliography of reports on the Fluorox moving-bed for producing UO3, UF4, and UF6 is given.

Departures from the usual precipitation method produced crystalline uranium peroxide in several forms. Three types of segregated needles were used in the preparation of three pellets for pellet-type elements. Densities of the pellets ranged from 93 to 97% of theoretical, depending on conditions of precipitation.

The ORNL analog computer was used to simulate four widely different experiment conditions in the small EGCR in-pile loops. Various control and safety systems were evaluated in each case. Curves are included which show the response to postulated accidents and component failures. The results indicate that one standard control and safety system will probably be adequate for a wide variety of test conditions.

Short reports to the members of the Nuclear Cross Section Advisory Group from three groups: (1) High voltage group; (2) Fast chopper time-of flight spectrometer; and (3) Electronuclear research division.

This report is presented in order to provide a satisfactory system of symbols and identifications for process-instrumentation equipment and to promote a uniformity of practice that will simplify and expedite instrumentation work. It is intended that the systems presented here should be capable of designating and identifying the multitude of instrumentation items which are used for control and operation of conventional processes, as well as for specialized work peculiar to ORNL. Instrument Society of America standards have been adhered to whenever practical.

Equipoise-2, a two-dimensional, two-group neutron-diffusion code in R-Z geometry, has been programmed for the IBM-7090 computer. This code was designed to permit the running of large numbers of cases without requiring excessive machine time. Typical running times are of the order of one to three minutes per case for a 1000-point problems. The maximum number of mesh points that can be used is 1444.

It is the purpose of this memo to clarify and revise burnup calculations for the APPR irradiation program and to present a proposed irradiation schedule. the report will also be in the nature of a review of the program.

In order to test the operation of a freeze jacket in air with the process fluid closely following the conditions found in the high pressure circulation loop of a homogeneous reactor, or, 2000 psi and 300C, a test loop was built and used in conjunction with existing refrigeration equipment. The freeze jacket was made of 5/16-in. type 346 stainless steel tubing wrapped around and welded to a 1/2-in. stainless steel process line. It was concluded that for these reactor operating conditions, only a small leak rate, 11 cc/min, could be frozen off. It is recommended that, at the beginning of the the freezing operation, the temperature of the secondary refrigerant entering the freeze jacket be at least -40C and that the freeze jacket be made as long as practical.

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 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.