Activity ratios for the important members of the U/sup 232/ chain are presented in tabular form. These ratios are presented for initially pure U/sup 232/, initially pure Th/sup 228/ and initially pure Ra/sup 224/.- (auth)

Processes for the conversion of uranium compounds or uranium metal to uranium hexafluoride ordinarily involve the use of a powerful fluorinating agent. Elemental fluorine is used when the scale of operations justifies the construction of a fluorine generating plant, but for smaller operation the use of the interhalogens of fluorine has definite advantages. These compounds provide a high concentration of fluorinating power at moderate temperatures and pressures and are more easily stored and transported than fluorine. In addition, fluorinations in the liquid phase often proceed more smoothly than those with gaseous fluorine. However, the use of . the interhalogens introduces the problem of separating the uranium hexafluoride from the unreacted reagent and from any by-products which may have been formed. The present work is concerned with the determination of the phase equilibrai among the materials uranium hexafluoride, chlorine trifluoride, and hydrogen fluoride. metal with chlorine trifluoride-hydrogen fluoride solutions or as a result of treating many uranium compounds and ores with chlorine trifluoride. These phase equilibria define the physical conditions necessary for separating the components by the processes of crystallization or distillation and have made possinle the successful Operation of a pilot plant for the direct recovery of uranium hexafluoride from …

A third thermoelectric generator has been constructed and tested. Specifications for batteries having outpnts of over six volts at power levels of 30, 50, and 1000 milliwatts at the end of six months have been made. An attempt was made to verify experimentally the theoretical equations of the Thomson and Peltier effects. (For preceding period see MLM-CF-58-1-40.) (auth)

The manufacturers' specifications for temperature coefficient of voltage and internaI impedance of the compensated avalanche diode types IN430A and IN430B appear quite promising. These devices couId be used as shunt regulators in high stability power supplies if the noise and drift rate were sufficiently small. One investigator reported the voltage did not drift more than the J-57 engin 0.002 per cent over a 7,000 hour period. Stability tests were performed on two diode samples under resonable laboratory conditions. The measured drift rate did not exceed 0.005 per cent per month, and short term noise was less than the J-57 engin 0.002 per cent. The actual diode drift rate may be even lower than the measured rate. (auth)

The U--Zr phase diagram is presented. Melting points of B and other oxides ahove 2000 deg C were determined. The structure of Ca and Sr aluminates was refined, and Al/sub 2/)/sub 3/UO/sub 2/ bodies were fabricated. Techniques for impregnating leached unfired Vycor are being standardized. The effects of radiation on hardness of Vycor and on Pu-- U oxides have been obtained. (C.H.)

The 5509 fuel rods for the Yankee Atomic Electric Company critical experiment were made by sealing sintered UO/sub 2/ pellets in stainless steel tubing by welding end plugs to each end of the steel tubes. The complete history of the fabrication of these fuel rods from the receipt of pellets to the radiographic inspection of the completed rods is discussed in detail. The control of the quality of the UO/sub 2/ powder was exercised by com paring certified analyses supplied by the vendor with analyses made by APD on correspondng lots. Chemical analyses, inspection data, procedures and processes are presented in tables and graphs. (auth)

Progress is reported in developmental studies on the Redox Process and studies on the chemistry of Ru. (C.H.)

Experimental results on core structural materials, moderator, and fuel materials are discussed. Data are given on grain growth of Zr and thermal conductivity, thermal expansion, and temperatures of various types of graphite. (M.H.R.)

Tests on 1-gpm and 0.5-gpm hydroclones showed them to be unnecessary. Solvent-aqueous phase separation, although effective when optimized, was infrequently required. The 1-gpm hydroclone produced better water separation than the 0.5-gpm hydroclone for similar control parameters. Decontamination factors for widely varying flow conditions ranged from 1 to 3. Introduction of water into the hydroclone feed stream had no effect on decontamination or particle removal. (auth)

The heat geueration and thermal stresses in spherical HRE-4 vessels 3 to 4 1/2 ft in diameter with clad and solid stainless steel walls were invegtigated. Parameters included thorium slurry concentrations and moderator material (D/sub 2/ O and H/sub 2/O). The prirnary purpose of this study was to determine the influence of thermal stresses on the selection of the core size for the HRE4 reactor. Curves are presented which facilitate relatively rapid determination of stresses for the range of vessels considered. It is concluded that steady- state thermal stresses in the clad or solid stainless steel vessels considered will not have to be a determining factor in the selection of a core size, provided the power density does not exceed approximately 15 kw/l in the clad vessels and 8 kw/l in the solid stainless steel vessels. (auth)

A general survey is presented of information in the uranium alloy field. Emphasis is centered on alpha uranium-rich alloys of special interest as uranium- base fuel element materials. The systems treated include uranium-molybdenum, uranium-silicon, uraniumzirconium, uranium-niobium, and uranium-zirconiumniobium (high uranium compositions). The uraniumfissium alloys are discussed in relation to their projected applications as fast reactor fuels. Equilibrium diagrams, transformation kinetics, and other pertinent data are presented for the uranium plus fissium'' element systems, e.g., uranium-ruthenium, uranium-rhodium, uraninum-palladium, and uraniummolybdenum-ruthenium. The properties covered include constitution phase diagrams, metallographic structure, physical and mechanicaly properties, transformation kinetics, crystallographic structure, thermal cycling, ad irradiation stability (where pertinent). Correlations between microstructure, heat treatment, and dimensional stability are emphasized, with particular reference to the U-2 wt.% Zr, U-5 wt.% Zr, and 1 1/2 wt.% Nb alloys. A discussion of the role of alloying and heattreatment in improving the dimensionaly stability and corrosion resistance of uraaium is presented, and an evaluation is made of the present status in attaining these objectives. (auth)

Applications of the various iodine isotopes in diagnosis and therapy are discussed. Problems of dosimetry and radiation exposure to patients and hospital personnel are considered. Some quantitative aspects of radiation damage in mammals relevant to the clinical use of radioiodine are reviewed. Emphasis is placed on applications of iodine-131 in diagnosis and therapy of hyperthyroidism and hypothyroidism; in the treatmeat of thyroid carcinoma; and in thyroid ablation for cardiac disease. 54 references. (C.H.)

>The Boltzmann equation for the neutron distribution in a reactor is discussed. Analytic and numerical solutions are presented. The general results are illustrated with data on the performance of the SPERT and KEWB tests. (W.D.M.)

. A review of recent dcvelopments and improvements in aqueous processes for accomplishing separation and decontamination of irradiated fuels from power reactors is presented Research and development is currently being pursued in tbe United States on three distinct types of fuel processing methods; pyrometallurgical processes, fluoride volatility processes, and aqueous processes. Although the ultimate role of these processing methods in a nuclear power economy cannot be accurately assessed at the present time, it is felt that the proven reliabilita and versatility of aqueous processes guarantees them a prominent role in power reactor fuel reprocessing. Aqueous solvent extraction processes, for example, are ideally suited for installation in central processing plants which are designed to handle fuels from a number of power reactors generating a total of several thousand megawatts or more of power. Under these circumstances, nuclear fuels can be processed by continuous processes at high throughputs and at high on-stream efficiency and therefore at low unit cost. (auth)

Progress is reported in the study of ThO/sub 2/ sols and related systems and in the development of methods for measuring the physical properties of sols and slurries. (W.J.H.)

Experimental reactor physics data for thorium water reactors are summarized. Reactivity lifettme considerations are presented including a comparison of the reactivity liietime of U/sup 238/ and Th reactors fueled with U/ sup 235/. Critical experiments and results on several water moderated enriched U- Th assemblies are reported. Values of infinite medium multiplication constants and migration areas were derived using the supplementary measurement of the rate of change of reactivity with water height as a function of water height. The effective resonance integral of the metal and ThO/sub 2/ in the water lattices were derived from measurements of cadmium ratios of Th. The experimental results were compared with theory. (M.H.R.)

Maximum thermal-neutron fluxes obtainable for the case that maximum power density is limited to 100 w-/cm/sup 3/, and for the case that total reactor power is 100 Mw are listed for a homogeneous reactor, and for five moderators. (auth)

The history of the safety device program and its accomplishments to date are described. Safety devices discussed are divided into those for production reactors, research reactors, and low power research reactors and zero power critical assemblies. Sketches of each type are included. (M.H.R.)

The purpose of this report is to present the revised process design to be used as the basis for the detail design of Redox dissolvers each of which shall meet the following: Process irradiated natural uranium slugs (standard Hanford or I&E type) at 43% of Phase II Redox production rates; process irradiated enriched uranium up to .94% U{sup 235} isotopic content in slugs (standard Hanford or I&E type) at rates up to 43% of Phase II capacity; and process irradiated fuel elements of cluster design from new reactors with cladding material of aluminum or zirconium, enrichments to .94% U{sup 235} and dimensions of ten feet long and 10 inches square. This report is also to form the basis for slug distribution tests to be performed by HLO concurrent with the final design effort.

The corrosion rate was determined and tabulated for stainless steels, nickel alloys, carbon steels, Zircaloy-2, and niobium in boiler water containing various concentrations of oxygen. The corrosion resistance of stainless steels and nickel alloys was compared with the resistance of the austenitic stainless steels in uranyl sulfate at 200 to 295 deg C. Additional loop runs with UO/sub 3/ --Li/sub 2/CO/sub 3/ --CO/sub 2/ so lution have confirmed previous data with regard to its nonaggressive nature to carbon steel. The corrosion of 347 stainless steel was investigated in 0.04M uranyl sulfate containing 0.04M beryllium sulfate and 0.006M copper sulfate, in 0.075M beryllium sulfate, and in 0.75M beryllium sulfate containing 0.25M dissolved uranium trioxide. centration to 0.03M or decreasing the uranium concentration to 0.02M in simulated HRT fuel solution (0.04M uranyl sulfate, 0.02M sulfuric acid, and 0.005M copper sulfate) has no significant effect on the solution stability or corrosiveness of the solution. The effect of varying the oxygen concentration between 250 and 2800 ppm in 0.17M uranyl sulfate at 250 deg C on the corrosion rate of 347 stainless steel was found to be negligible in the absence of added Cr/sup 4+/. The corrosion rate of 347 stainless steel in dilute …

6 6 5 4 ; 6 6 5 6 4 8 4 : the chemica: stability of possible fuel solutions, and the results obtained from these runs have been compared with data obtained by others. A solution containing 0.04 m UO/sub 2/SO/ sub 4/, 0.03 m CuSO/sub 4/, 0.03 m NiSO/sub 4/, and 0.023 m D/sub 2/SO/sub 4/ in heavy water was not completely stable at 175 deg C or at other higher temperatures. About 7% of the copper was lost at 175 to 275 deg C, and an additional 9% was lost when the temperature was increased to 300 deg C. In the same solution the nickel solubility decreased by 40% up to 200 deg C and then increased up to 275 deg C and at 300 deg C again decreased. Throughout the run no uranium was lost from solution. A second solution containing 0.01 m UO/sub 2/ SO/sub 4/, 0.02 m D/sub 2/SO/sub 4/, 0.01 m CuSO/sub 4/ and 850 ppm nic kel in heavy water was stable at 300 deg C. However, when an additional 450 ppm of nickel was added to the solution, about 10% of the uranium, 6% of the copper, and 20% more nickel …

For presentation at the American Chemical Society's Nuclear Technology Symposium, Boston, Apr. 5-10, Information is presented from studies with materials and conditions simulating those expected in anm aqueous homogeneous reactor for the formation of insoluble corrosion and fission products in aqueous UO/sub 2/ SO/sub 4/ solutions used as a reactor fuel. Sulfate salts which become insoluble as the temperature inereases deposited preferentially and reversibly on hot walls. Oxides from hydrolysis of sulfate salts deposit on any surfaces while insoluble corrosion products released into the solution show variable degrees of circulation. (auth)

Data are given on properties of three sets of cylindrical Be reflected, graphite-mcderated critical assemblies. The first set was primarily to establish characteristics as functions of C/Oy atomic ratio of core with nearly constant reflector thickness. Fission distributions were determined. The second set consisted of three assemblies with fixed core to determine the effect of redistributing reflector from the ends to the cylindrical wall. This series was done to provide the Los Alamos Scientific Laboratory's Theoretical Division checks for two-dimensional diffusion code. Flux distributions in the uniformlyreflected assembly were mapped extensively with bare and Cd-shielded foils of oralloy, Au, and In. The third set was to establish the minimum-volume core at C/Oy ~ 350 that could be made critical with available Be. Experimental critical data converted to equivalent spherical systems are compared with resuIts of S/sub 4/ calculations. (auth)

Irradiation by fast particles changes most physical and chemical properties of graphite by producing various lattice defects such as displaced atoms, vacancies, dislocations, etc. Attempts have been made to evaluate the contributions of different types of defects to the property changes. Such evaluations have proven rather successful for physical properties and will be extended to several chemical properties. Experiments are described which show that oxidation catalysts and inhibitors act differently on irradiated and on unirradiated crystals. The oxidation of graphite crystals in ozone was studied because it appears to provide clues to the enhancement of oxidation rates in the presence of ionizing radiation. The kinetics and mode of attack of -ozone on graphite crystals with or without prior neutron irradiation are described. (auth)