Results are given from studies on: mechanical properties of Zircaloy-2; preparation of Al--U alloys; effects of N, Ca, Zr, and Mo on electrical properties of UO/sub 2/; effect of chloride, fluoride, and sulfate contamination on corrosion of type 304ELC stainless steel in boiling HNO/sub 3/; density of Mallinckrodt slag and MgF/sub 2/-base materials; development of gamma-phase U alloys; preparation of U--Nb alloys; oxidation of Nb and Nb alloys in dry air at 1000 and 1200 deg C; thermal expansion of UBe/sub 13/ and UC; phases and compositions in C --N--U system; creep of arc-cast Ta during exposure to high- temperature Nas pressure bonding of PWR fuel platess evaluation of UC as reactor fuels and fabrication of UO/sub 2/ fuel pellets for NMSR loop tests. (For preceding period see BMI-1259.) (T.R.H.)

5 8 5 of 15% cold-worked Zircaloy-2 as material for process tubes for the NPR has indicated the need for study of the creep properties at elevated temperatures. Single crystals of high-purity Mo are being tify factors affecting irradiation-induced volume changes in graphite by means of sink-float density measurements was continued. The simulation of conditions after lossof- coolant incident in the PRTR by means of a digital computer was continued. The study of the radiometric analysis of Mg was resumed. The program on radiationinduced graft polymerization ss continuing with emphasis on the investigation of factors influencing the formation of free-radical and peroxy grafting sites. A study of the effect of gamma radiation on nitration reactions in the HNO/sub 3/ -cyclohexane system is reported. The manner in which U metal solidifies in cylindrical graphite molds is being studied. A study to determine the feasibility of coating Ni and Ni alloys with Ca by the arcspraying method is presented. A project to investigate the effects of Y/sub 2/O/sub 3/, La/sub 2/O/ sub 3/, and CaO additions to UO/sub 2/ with regard to stabilizing the UO/sub 2/ fluorite structure under both oxidizing and reducing conditions is presented. An evaluation is being made of …

Measurements of thermal conductivity and electrical resistivity were continued on irradiated specimens of UO/sub 2/-clad in Zircaloy-2 with NaK as the heat-transfer medium. The creep strength of 0.15% cold-worked Zircaloy-2 is being determined at elevated temperatures. Work continued on the development of highstrength corrosion-resistant Zr alloys. Research to develop a method of sink- float density measarements to identify factors affecting irradiation-induced volume changes in graphite was continued. Single crystals of high-purity Mo are being prepared for irradiation damage studies. Casting methods are presented for Al--U alloys. Time-temperature relationships are being studied during the solidification of small cylindrical U castings to improve the quality of fabricated U fuel slugs. Methods for preparing a thin protective coating of Mo on Croloy are being investigated. Research indicates that in a binary system of UO/sub 2/-- La/sub 2/O/sub 3/ 60 wt.% La/sub 2/O/sub 3/ is required to form a stoichiometric structure after oxidation. Investigations of the hydrides of U- Zr alloy for possible use in gas-cooled reactors were continued. An irradiation- damage program on type 347 stainless steel was undertaken to study the changes in physical properties that are caused by continued exposure to a fast-neutron flux up to 3 x 10/sup 15/ nvt …

Data are presented on the creep properties of 15% coldworked Zircaloy-2 at 290, 345, and 400 deg C. The progress on the development of an isotopic- exchange fuel-element leak-detection system is summarized. A program to develop a thermal-neutron-flux monitoring system for Hanford reactors is reported. Work on the radiometric analysis of calcium in cement was continued. A surveillance program is in progress to determine the effects of fastneutron irradiation on the mechanical properties of AISI Type 347 stainless steel. A summary of corrosion results obtained on Nb, Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti-Cr, and Nb- Ti-V alloys exposed in hightemperature water and steam is presented. A study of the creep characteristics of Zircaloy-2 at elevated temperatures during exposure to a fast-neutron flux was initiated. A research program was initiated to develop an analytical technique for monitoring the oxygen concentration in largescale radioactive sodium systems. Data are presented on the corrosion behavior cf Nb-U alloys in high-temperature water after 98 days. The development of Th -U-base alloys of improved radiation stability and corrosion resistance is reported. Methods of producing cermets of 90% of theoretical density or better containing 60 to 90 vol.% ceramic fuel are being investigated. The gas-pressure- bonding …

Data are given on: creep propenties of 15% coldworked Zircaloy-2; the liquidus in Al-U; effect of chloride on corrosion of type 304 ELC stainless steel; corrosion testing of stainless steels, Ti 75A lead, and rigid PVC as container materials for HNO/sub 3/-recovery process; viscosities of MgF/sub 2/ and MgF/sub 2/-MgO slags; phase boundaries in the U-Zr-H system; oxidation of Nb, Nb-W, and Nb-V alloys in dry air at 1000 and 1200 deg C; nonoxidizing properties of UO/sub 2/- La/sub 2/O/sub 3/ and UO/sub 2/-Sc/sub 2/O/sub 3/; study of bounding fuedamentals; hydrogen adsorotion by Nb; diffusion of H in ZrH/sub 4/; reflector controlled heterogeneous boiling reactor; pressure bonding of Zircaloy- 2-clad compartmented UO/sub 2/ fuel plates for PWR; and evaluation of UC as reactor fuel. (For preceding period see BMI-1247.) (T.R.H.)

Thermal-conductivity measurements are in progress on an unirradiated, unclad, natural U specimen. Data are presented on thermal conductivity measurements performed on UO/sub 2/. The creep properties of annealed and of 15% cold-worked Zircaloy-2 are being studied. A program was initiated to evaluate loss-of-coolant incidents in the PRTR by means of simulation on a digital computer. Research on the casting of hollow Al-35 wt. extrusion billets is reported. Further refinement of the method developed for the analysis of Mg in cement is in progress. The infrared and gaschromatography analysis of irradiated dodecane, decane, cetane. and octane, and their urea complexes, were continued. The manner in which U metal solidifies in cylindrical graphite molds is under study. Work has continued on development of a stabilized hightemperature nuclear fuel capable of operation in either oxidizing or reducing atmospheres. Progress in the stud of potential fueled moderators has continued with the determination of hydrogen-absorption isotherms for the Zr-25 wt. alloy. The effect of fast-neutron flux on the mechanical properties of AISI Tvpe 347 stainless steel are being determined and evaluated. The forging of Nb-U alloys is reported. Thorium-uranium alloys are being studied for the purpose of developing improved corrosion resistance and irradiation stability of …

Investigations are reported in: creep properties of annealed Zircaloy-2 and -3 at high temperatures; burst strength of welded Zircaloy-2 tubes; air oxidation of Nb, Nb-V, Nb-Mo, and Nb-Zr at 1000 deg C; vapor deposition of Mo coatings on stainless steel tubing; bonding fundamentals; niobium-hydrogen reactions; creep and stressrupture of sintered Ta at 1200 deg F in He; and determination of O in liquid Na by gettering. (For preceding period see BMI- 1201.) (T.R.H.)

A methed is descrihed for the treatment of radioactive neutralized nitric acid waste solutions. The process consists of electrolysis of sodium nitrate solutions and recycle the caustic formed to neutralize incoming nitric acid ate. Removal of a large fraction of the bulk inert substances by this method allows greater reductions in waste volume than conventional treatment by evaporation alone. Process chemistry studies show that a large fraction of the nitrate ion in alkaline solution can be electrolytically reduced to gaseous nitrogen reduction products at reasonable current efficiencies. Process engineering studies have resulted in the development of a prediction equation for the design of electrolysis cells used in the process. Pilot plant studies using cold'' nitric acid waste have demonstrated the feasibility of the electrolytic recycle methed. A preliminary cost estimate was performed which iadicates a cost of from 75 to 95 per initial gallon of waste processed. Basis for this estimate is a conceptual design of a plant capable of handling the highlevel waste produced in reprocessing one ton of uranium metal per day. These costs are considered reasonable in a nuclear power economy, and indications are that ultimate disposal of the concentrated waste in clay may be possible. (auth)

The effect of initial particle size (6.2 and 14.7 mu ), forming pressure (50,000 and 100,000 psi), temperature (1600 and 1800 deg C), and time at temperature (1 and 3 hr) on the bulk density, grain size, and porosity of sintered uranium monocarbide compacts was studied, Within the limits of the variables studied there appeared to be no significant changes in these properties of the sintered material attributable to the changes in the fabricating conditions. (auth)

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The refractory uranium compounds UC and UC/sub 2/ were prepared by arc- melting methods, UN by gas solid reactions, and UB/sub 2/ and UBe/aub 13/ by solid-solid reactions. Techniques for sintering powder compacts of these compounds to densities 90% of theoretical or greater were developed. Mean linear thermal-expansion coefficients of 7.9 x 10/sup -6/ per deg F for Uc and 9.3 x 1O/ sup -6/ per deg F for UBe/sub 13/ were obtained in the range 68 to 1800 deg F. The thermal conductivity of sintered UBe/sub 13/ increased from 0.050 cal/ (sec)(cm)(C) at 1OO deg c to 0.064 cal/(sec)(cm)(C) at 650 deg C. The electrical resistivity of sintered UBe/sub 13/ was 113 microhm-cm at 27.6 deg c. Knoop microhardness values varied from 500 for sintered UC/sub 2/ to 1155 for sintered UBe/sub 13/. Sintered Uc and Uc/sub 2/ decompose rapidly in boiling (atmospheric pressure) water. A sintered UBe/sub 13/ compact lost 2.2 mg per cm/sup 2/ (average), after 72 hr in boiling water (atmospheric pressure). Uc/sub 2/ reacts more rapidly with nitrogen and oxygen than does pure uranium. UB/sub 2/ and UBe/ sub 13/ were less reactive with oxygen and nitrogen than is pure uranium. The heat of formation …