Potential cladding materials for a flat-plate fuel element containing a dispersion of UC or UC/sub 2/ in U--15 wt.% Mo alloy were surveyed on the bases of compatibility with the fissile compounds, matrix material, protective cover materials, and liquid sodium as well as the feasibility of fabricating fuel plates by roll cladding. Radiative-capture cross sections, thermodynamic data, eutectic and intermediate compound formation, mechanical properties, and corrosion by 1000 tained F Na are reported for austenitic stainless steels, chromium, nickel, niobium, molybdenum, tantalum, vanadium, and zirconium. It was recommended that "A" nickel (molybdenum barrier), Zr-3 wt.% Al. Nb--2 wt.% Cr, and Fansteel 82 be relected for investigation. (auth)

None

Current results indicate probable over-all rates of about 0.2 mils/month for titanium vs. 1.5 to 3.0 mils/month for Ni-o-nel. Tests are not 100% comparable due to changes made in flowsheet conditions, but they have been of sufficient variation and length as to allow good predictions to be made. Both metals show some tendency toward local attack in Thorex solutions. These tendencies are increased by poor welding techniques. There is a possibility that a Ni-onel dissolver could be used for interim Zirflex processing although the rates obtained in scouting tests are relatively high (6 to 8 mils/month). Titanium shows no promise of usefulness in either Zirflex or Sulfex on the basis of scouting tests. It can be used for dissolution of U-Al fuels in 8M HNO/sub 3/- Hg(N0/sub 3/)/sub 2/ (rate < 1 mil/month). If the choice between these two processes were made on the basis of present corrosion results alone, Darex-Thorex would be chosen. (auth)

Further studies are reported on the analytical behavior and experimental testing of Cs/sup 137/ plaque irradiator designs. Low-level sources used for the initial experiments consisted of about 7 mc of Cs/sup 137/ aqueous solution in brass trays 20 by 20 by 1/2 in. high having wall thicknesses of 1/16 in. Calibration tests were made preliminary to radiation field mapping. (T.R. H.)

The concept of operating bearings and gears unlubricated in a helium environment was questioned as to fundamental soundness. Tests were made to determine if immediate gross failure occurs. A bearing of standard manufacture was operated in helium for 42 hr. A high degree of gradual wear was encountered. No gross or sudden failures were found. (W.L.H.)

This document details activities of the Fuels Preparation Department during the month of December 1959. (FI)

This document details activities of the Fuels Preparation Department during the month of June 1960. (FI)

This document details activities of the Fuels Preparation Department during the month of March 1960. (FI)

A calculation was made to determine the fuel plate heat fluxes resulting from after shutdown fission product heating. Fission product source strengths were obtained via the IBM Internuc code. Slab geometry was assumed. The results indicated that the maximum heat flux would occur slightly inboard of the center of the fuel annulus, with the heat flux at the inner annulus radius running about 8% below the maximum, and the outer radius heat flux 10% below the maximum. For decay times of 1.0, 10, 10/sup 2/, 10/sup 3/, 10/sup 4/, and 10/sup 5/ seconds the maximum calculated fuel plate heat fluxes were 42.0, 30.0, 18.0, 9.5, 4.0, and 1.3 x 10/sup 3/ Btu/hr-ft/sup 2/, respectively. The core coolant gamma heating rate during reactor operation was also calculated using the same techniques, but including the fission and capture gamma sources. Average coolant gamma heat generation rate was about 33 watts/cc at the start of the fuel cycle, and 57 watts/cc after the fission products built up. (auth)

An estimate is made of the illumination required for the 80-in. bubble chamber, based on the absolute sensitivity of the film. The required intensity is estimated to be 23.4 tube is calculated to be 145 cm/sup 2/. The results are compared with these obtained from 72- and 20-inch chambers. (D.L.C.)

Objective of the test was to determine and evaluate the behavior of 8 20-inch fuel elements during irradiation using the hot-headed end closures on the inner tube.

The advantages of the use of organic liquids in fission reactors to minmize corrosion and pressure problems were studied relative to the solution of thorium in such fluids. Thorio-organic compounds were prepared from organic acids, diketones, and other chelating compounds. Salts of carboxylic and phospho- organic acids were insoluble. The chelate with dibenzoylmethane was soluble in molten biphenyl but was decomposed at 300 deg C. The general low solubility of thorio-organic compounds in nonpolar solvents can be explained by steric effects. The large thorium atom has the ability to form strong coordination complexes with adjacent molecules, leading to coordination polymers. The effect can be minimized by shielding the thorium nucleus with large organic groups such as dibenzoylmethane. The large, branched organic groups needed to impart solubility limit the maximum solubility. (auth)

A test was made to determine the direction and magnitude of any drift in the temperature-sensing elements and the receiver-indicating units for the reactor coolknt loop and pressurizer resistance thermometers. A test was also made to determine the difference between the measured value of the neutron flux with the reactor shut down, as determined by BF/sub 3/ counters, and the neutron flux due to the Po-Be source. (W.L.H.)

Test quantities of thoria (aporoximately 3800 lb of thorium oxide and 3200 lb of mixed thorium-uranium oxide) were prepared during FY-1960 for members of the Reactor Experimental Engineering Division (REED). For preparation of thorium oxide, the calcination temperature was varied from 650 to 1800 deg C. The surface area of the fired oxide ranged from 0.5 to 25 m/sup 2//g and the mean particle size ranged from 1 to 8 microns. For preparation of mixed thorium-- uranium oxide, the calcination temperature was varied from 1050 to 1225 deg C. The mixed oxide contained either 0.5, 3.0, or 8.0 wt.% of uranium as requested with a mean particle size of 1.5 microns. The over-all losses of thorium and uranium were 19.4 and 18.3%, re spectively. Four 1400 lb batches of thorium nitrate were dissolved to prepare solutions containing 300 to 350 g/liter of thorium for solvent extraction studies in the Chemical Technolcgy Division and 2100 lb of dry, recovered waste thorium oxide was supplied to the ORNL Fuel Cycle Program. (auth)

The KER-1 Zircaloy process tube extracted from the reactor in May 1960, is currently being examined in Radiometallurgy with particular attention to an area of heavy oxidation and hydriding revealed during a burst test. The area in question is approximately 1 1/4in. wide and extends longitudinally an unknown distance from the points where sections have been taken. The oxide layer of interest is about 8 mils thick at the center, thinning at the edges. The ``oxide`` is quite adherent with only relatively minor systems of longitudinal cracks observed. There is no corresponding thick oxide on the outer surface of the tube. The hydride region decreases from an estimated 1000 to 2000 ppM at the oxide-metal interface to about 30 ppM at the outer surface. Assuming a linear concentration gradient, the average hydrogen content in the affected zone might be about 750 ppM (Examination of the metal at the midpoint of the tube section suggests 750 ppM is-probably excessive). The region of high hydride concentration extends circumferentially past the heavily ``oxidized`` region, presumably because both tube temperature and hydride concentration gradients are favorable to hydrogen diffusion.

A possible phase transition in liquid He{sup 3} has been investigated theoretically by generalizing the Bardeen, Cooper, and Schrieffer equations for the transition temperature in the manner suggested by Cooper, Mills, and Sessler. The equations are transformed into a form suitable for numerical solution and an expression is given for the transition temperature at which liquid He{sup 3} will change to highly correlated phase. Following a suggestion of Hottelson, it is shown that the phase transition is a consequence of the interaction of particles in relative D-states. The predicted value of the transition temperature depends on the assumed form of the effective single-particle potential and the interaction between He{sup 3} atoms. The most important aspects of the single-particle potential are related to the thermodynamic properties of the liquid just above the transition temperature. Two choices of the two-particle interaction, oonstituent with experiments, yield a second-order transition at a temperature between approximately 0.01 K and 0.1 K. The highly correlated phase should exhibit enhanced fluidity.

A program consisting of two related phases is described Ia Phase I, a study was made to find the relationship between the amount and nature of radioactivity, particle size distribution, and weight of particulate matter present in the lower troposphere. Emphasis was placed on the distribu tion of strontium-90 and total beta activity. Results of a limited number of analyses indicate that strontium-90 and total beta activity is associated primarily with particles below approximately 0.1 micron diameter. Phase II consisted of experimental studies on scavenging of solid particulate matter by water droplets. It was found that water vapor gradient around a condensing droplet promotes scavenging of particles of 1.3 micron and 0.3 micron diameters. The effect of water vapor gradient around an evaporating droplet is not well defined. (auth)

The effects of ternary additions of up to 3 wt.% Sn or Zr to an Al-35 wt.% U extrusion alloy were evaluated on the basis of casting characteristics, UAl/sub 3/ retention, extrusion behavior, mechanical properties, and corrosion resistance. Both additions increased the fluidity of the alloy, and both promoted retention of UAl/sub 3/. The best fluidity was obtained by a 2 wt.% Sn addition, while Zr was the more effective stabilizer of UAl/sub 3/. The retention of UAl/sub 3/ decreased the extrusion pressure needed for fabrication and caused a corresponding decrease in tensile and creep-rupture properties. Reductions in strength were most noticeable at elevated temperatures. The 1000- hr stress-rupture strength of the binary alloy at 200 deg C (8300 psi) was approximately 25 and 11% higher, respectively, than the alloys containing 3 wt.% tin (6200 psi and 3 wt.% zirconium (7400 psi). The additions either slightly improved or had no effect upon the resistance of the Al-35 wt.% alloy in 150 deg C demineralized water. (auth)

Cobalt samples frequently are irradiated in nuclear reactors to produce gamma sources and can be irradiated as integral flux monitors because of the long half-life of the isotope produced. At the present time a small cobalt sample is being irradiated within the KW Reactor Snout facility for future use as a radiographic source for inspection of finished product in the Chemical Processing Department. Analysis was made to estimate the buildup of activity in this sample; the general equation may be of interest and value for other cobalt sample irradiations.

The objective of this test detailed in this report is to irradiate a test fuel assembly for the MGCR (Maritime Gas Cooled Reactor). The irradiation of this assembly will be carried out in the DR-1 Loop under controlled conditions to determine the feasibility of the heterogeneous 19-rod bundle fuel element concept. Diffusion of fission products through metal cladding. Fission gas retention of fuel bodies. Dimensional stability of fuel bodies, and satisfactory performance of the creep-shrink process for maintaining pellet position in the fuel pin.

Changes in operation and corresponding changes in the reactivity status of Hanford reactors are the result of a continuing effort to improve operating efficiency. Trends data related to these changes in operation and reactivity have been published previously for the periods from 1950 through 1958. The purpose of this report is to include trends data for 1959. Bar graphs in the first part of the report show yearly averages of selected data, and tables in the last part of the report show maximum, average, and minimum values. This document presents trends data for B, D, DR, and F reactors while a second document, HW-64932, presents trends data for C, H, KE, and KW reactors. Data included in past years which have not been included in this report are trends in pile power level at shutdown omitted due to a security status change regarding power levels, and number of temporary poison columns per startup omitted due to virtual elimination of temporary poison startups at B, D, DR, and F Reactors; added were potential non-equilibrium gains and potential equilibrium gains. Notice that all reactivity values are listed in the unit per cent excess k.

Tables present excess reactivity vs pile rising periods, excess reactivity vs pile falling periods, mint-nominal reactivity vs charge length, Pb-Cd nominal reactivity vs charge length, and nominal reactivity of poison splines.

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The work carried out over a 2 1/2-yr period on the scavenging of radioactive particles which might be released by the reactor system of a nuclear- powered ship is summarized. Two types of dispersions were considered: aerosols and hydrosols. Radioactive aerosols were scavenged by heterogeneous coagulation with solid and liquid aerosols produced within the radioactive aerosol cloud. Liquid or highly hygroscopic particles, which can be classified as solid particles with liquld films on their surfaces, were found to be the most effective scavengers. A system of fine water spray and hydrolysis products of silicon tetrafluoride was found to be suitable for field application. Scavenging of radioactive cations, anions, and colloids of corrosion and fission products was studied in substitute ocean water, natural ocean water, and natural harbor water. A scavenging system composed of KMnO/sub 4/ and ferrous salts successfully removed most of the radioisotopes. Fe(OH)/sub 3/--MnO/sub 2/ hydrate adsorbed and absorbed radioactive species, thus transferring them from a liquid to a solid phase. Addition of Floc 111 to the system improved sedimentation. The KMnO/sub 4/-FeSO/sub 4/-Floc 111 system was found to bs suitable for field application. (auth)