SNAP 1 is the designation for the 0.5-kw nuclear auxiliary power supply intended for application in a satellite. SNAP 1 was designed to convert thermal energy from the decay of a radioisotope into electrical energy using a Rankine engine with mercury as the working fluid. A successful 2500-hour endurance test is described of a complete developmental version of the SNAP 1 power conversion system utilizing a prototype turbomachinery package, an electrically heated boiler, and an air-cooled condenser. Indications from the data obtained during the test and from inspection of the system following the test were that many more hours of satisfactory operation could have been obtained on all major system components except the rotating unit pump. The mercury-lubricated bearings, the turbine, and the alternator, all demonstrated excellent endurance capability. Based on previous component tests, it is concluded that the pump performance deterioration was caused by air entrainment in the liquid Hg. (auth)

A graphical compilation is presented of atomic and molecular cross sections of interest to controlled thermonuclear research. The cross sections are shown, as a function of energy, for collision processes involving molecular ion dissociation, charge exchange, excitation, ionization, photoionization, scattering, energy loss, and recombination. Pertinent nuclear cross sections are also included. A bibliography is given covering the literature since 1950. (auth)

SNAP II is the designation for a 3 kw nuclear auxiliary power unit to be used in a satellite vehicle. The SNAP II System consists of a reactor heat source, a boiler, a Hg Rankine engine, an alternator, and a condenser. The corrosion and subsequent mass transfer resulting from the use of Hg as the thermodynamic wo:king fluid are important considerations in the selection of materials for the SNAP II System. Consequently, corrosion and mass transfer behavior were under study for the past three years. Recent results of this study are presented and the corrosion mechanisms involved are discussed. (auth)

The Title I design fuel assembly and each of its components for the Experimental Gas-cooled Reactor are described. The development of the specific nondestructive testing techniques for the evaluation of the components and the fabricated fuel capsule is presented. These techniques include penetrants, pulse- echo and resonance ultrasonics, radiography, eddy-currents, helium leak testing, and others. Discussion is made on the capabilities and limitations of these techniques for the specific inspection problems and reasons presented for the selection of the inspection requirements. (auth)

The computer codes CROCK and SHOCK, written to aid power system designers in selecting the minimum weight point for space power plant radiators, are described. CROCK can be used to select the minimum weight design for a system in which the waste cycle heat is radiated directly to space from the condenser, while SHOCK can be used for systems in which the sensible heat lost from a single-phase fluid is radiated. (J.R.D.)

Development activities on the large-scale, hightemperature, thoria slurry facility at ORNL were suspended in July 1960. The latest design philosophy, the status of component development, operational procedures, and limitations of each section of the system are presented. The design of the facility would be adequate for a singleregion slurry-reactor experiment on a slurry blanket system with a properly developed blanket flow pattern. The equipment consists of a 330-liter circulating system, including a 32-in.-ID pressure vessel designed for service at 2000 psi and 300 deg C, a 3000-lb- capacity low-pressure system, and the interconnecting lines. Flow sheets are included and materials of construction and recommended instrumentation are discussed. The first five experimental runs in the system, which were made to determine the feasibility of operating with fertile thoria slurry in the blanket region of Homogeneous Reactor Experiment No. 2, either as-built or after minor modification of the original system are summarized. The first run, a heat exchanger test run made at 1800 psi and 300 deg C, was considered inconclusive. Repeated plugging was experienced, but this occurred in the connecting lines rather than in the heat exchanger. The next four runs were investigations of the high-pressure blanket system, with the pressure …

The resistsnce of the water --tributyl phosphate (TBP) interface to diffusion of uranyl nitrate was investigated by a photographic photometric technique. The resistance was below the limit of detectability in the experiments, even with a surface-active agent present in the system. The sensitivity of the experiments established an upper bound for the interfacial resistance to diffusion equivalent to that of a 25- mu film of TBP under steady- state diffusion. The results indicated that the rate-limiting step in solvent extraction of uranyl nitrate from water by TBP is transfer of the diffusing species between the interface and the bulk phase rather than across the interface. In the experiments a steady-state concentration profile was established across an interface by providing a source of uranyl nitrate on one side and a sink on the other in a special cell. Uranyl nitrate diffused along the concentration gradient and hence across the interface. A resistance to diffusion by the interface would be manifest as a departure from equilibrium of the uranyl concentrations immediately adjacent to and on either side of the interface. Accurate values of the interfacial concentrations at steady state were obtained by extrapolating the profile obtained from a few point measurements on …

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The preliminary design is described of equipment to carry out experiments on boiling burnout in which the average coolant density as a function of coolant channel length as well as local coolant densities may be measured. It appears that by use of the equipment, average densities can be accomplished in a few seconds, while determination of a complete density map (with 5% error) across a plane of the assembly requires approximately one hour. (J.R.D.)

Temperature profiles in stored irradiated LAMPRE (Los Alamos Molten Plutonium Reactor Experiment) fuel capsules were simulated both mathematically and experimentally. Each capsule which contains 90 at.% Pu and 10 at.% Fe fuel could produce up to 200 watts of fission product heat power after long term exposure in the reactor and could develop temperatures in storage that would accelerate the corrosion of the tantalum capsules. The experimental and mathematical simulations were made to investigate the temperature problems that could arise on storage of the capsules. The mathematical analysis produced solution equations from the four coupled secondorder differential equations by an iterative generation procedure. The generated equations were then used in an iterative computer program which produced the convergent temperature profiles by a calcuiation and correction technique. The iteratively generated equations, coupled with the computer program, produced temperatures which agreed very satisfactorily with the experimental simulation results. Because of this agreement, the iterative equation generation procedure can be recommended for similar complex longitudinal-radial heat transfer problems. (auth)

In support of the High Flux Isotope Reactor program, experimental determinations were made of friction factors, burnout heat fluxes, and average and local nonboiling heattransfer coefficients for forced-convection flow of water through this aluminum and nickel rectangular channels under the following conditions: heat flux = 0.1 x 10/sup 6/ to 7.4 x 10/sup 6/ Btu/hr - ft/sup 2/, velocity = 10 to 85 fps, Reynolds number = 9,000 to 270,000, pressure = 1 to 39 atmospheres absolute, flow gap = 0.043 to 0.057 in., and heated length = 12 and 18 in. A few tests were made to ascertain the effect of an axially oriented cylindrical spacer strip on surface temperature distribution and burnout heat flux. The results of these studies, are in reasonably good agreement with accepted ccrrelations. The friction factors are in satisfactory agreement with the Moody chart for the relative roughness of the test sections used, the burnout heat fluxes are well reproduced by the Soviet Zenkevich-Subbotin correlation, and the local and average heat-transfer coefficients are slightly larger than values predicted by the Hausen and Sieder-Tate equations. Miscellaneous experimental and analytical HFIR heat-transfer studies are included. (auth)

A full scale 15 deg sector of the P122 reactor configuration was constructed. The model was complete with respect to all internal cooling air passages, and reflectors, thermal shielding, and island reflector. The contract was terminated before any test data could be obtained. Investigation of the effect of atmospheric dust on performance of reactor systems using wire screen matrix fuel elements is reported. The interim conclusion is that dust would not limit aircraft performance or life. Work proposed but not completed is outlined. Appendices contain previously unpublished reports. (auth)

Reactivity effects of 93.2% enriched oralloy, molybdenum, iron, niobium, 310 stainless steel, 304 stainless steel, Inconel X, FeC rAl, zirconium, nickel, nichrome V, and chromium clad with FeCrAl, in the form of foils in the center cell of the BEM-II B minimum void, beryllium moderated critical experiment are reported. Nineteen-energy-level diffusion calculations with cell corrections from both coarse and fine energy detail were correlated with the measurements. The fine energy detail improved the correlation, mainly because resonance self- shielding and flux depression are more adequately treated, and residual discrepancies for the most part can be attributed to inadequacies in the cross section data. (auth)

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A method is presented for the aerothermodynamic optimization of the net power and/or propulsive thrust per unit reactor free flow area of a nuclear power plant operating on the Brayton cycle. A system so optimized will translate into the minimum size, therefore the minimum weight, nuclear system for any selection of reactor materials, lifetime, and fuel loading. The theory and development of the thermodynamic optimization process, the importance and effect of various parameters, and specific methods to be employed in the optimization of the various forms of the Brayton cycle are discussed. A sample calculation for the case of the ramjet application is included. The results of the application of these techniques to any Brayton cycle system may be used in conjunction with nuclear sizing methods, for beryllia-moderated reactors, to determine the required reactor size as a function of fuel loading and reactivity requirements. (auth)

A bibliography of eight general reviews on the chemistry of Po is presented. The characteristics of Po isotopes are tabulated. The oxidation states, compounds, complexes, and analytical chemistry of Po are reviewed. Radiochemical procedures for Po are presented. (D.L.C.)

Safety precautions utilized for control and cominement of fissionable and radioactive materials in the various aqueous reprocessing operations performed at the Idaho Chemical Processing Plant are presented. Three primary nuclear safety controls, geometrical, mass limitation, and concentration control, are used. Operations are penformed according to standard operating procedures which are set up to prevent circumvention of the primary nuclear safety controls. The various processing operations with their particular safety features are discussed. The operations include receipt, handling, and storage of irradiated fuel elements, dissolution of the fuel elements in various reagents, separation of the unburned fissionable material from fission products and fuel element structural materials by solvent extraction, salvage or recycle operations of off- specifications product or waste solutions that exceed the dispossble fuel concentrationalimits, product packaging, storage and shipment, fission product recovery, and waste collection, handling and disposal. The originai plant design and later additions and modifications included built-in geometrical control wherever practical with allowances for possible neutron interaction between vessels. The standard operating procedures specificaily state mass limits and concentration controls required for certain operations which involve appreciable quantities of uranium. Administrative control insures compliance with the standard operating procedures. (auth)

The Bevatron operated an average of 80% of the scheduled operating time with the exception of one 5-day shutdown period for emergency maintenance of the motor generator sets. The machine was shut off l9% of the scheduled operating time because of component failure and 1% at the request of the user. During and following the Christmas shutdown period an ultrasonic inspection was made of the flywheel keyway on each of the motor generator sets. No faults were apparent. Charpy samples were taken from the inspection bore in the flywheels. Subsequent"vee notch" Charpy impact tests indicated that the flywheels are operated in a brittle region rather than in the ductile range expected. The research program again was devoted mainly to study of pi - and K-meson interactions. (auth)

Isomeric cross section ratios for several nuclear reactions in the medium mass region were measured. The cross section ratios of isomers produced by the thermal neutron irradiation of Se/sup 60/, Rh/sup 103/, Pd/sup 106/ TE/sup 128/ and Cs/s up 133/ were measured. Isomeric cross section ratios for the formation of the isomeric pair In/sup 110/ and In/sup 110m/, produced in the Ag/ sup 107/ ( alpha ,n) and the Ag/sup 109/ ( alpha ,3n) reactions were also measured at several energies of the incoming projectile. He ions were accelerated to a maximum energy of 43 Mev. The relative amounts of the isomers produced in the reactions were determined by scintillation counting. In the case of the In/sup 110/ isomers the efficiencies of the scintillation counter for detecting each isomer were measured, so that absolute cross sections as well as cross section ratios could be determined for these isomers. The isomeric cross section ratios were also calculated theoretically using a statistical model, in which angular momentum effects were included. By comparing experimentally measured cross section ratios for the formation of isomeric pairs to the calculated ratios, information about the dependence of the nuclear level density on angular momentum was obtained. (auth)

Molecular sieves are used as the basic adsorber in a close-capture air recirculation system designed primarily for remote operation with master-siave equipment. A compact evaporator-dissolver unit provides a vessel for dissolution of an Al slug containing the radioactive material and provides an evaporator head under which volume reduction may be carried out within a specially prepared centrifuge cone. One movable condenser serves both operations. Resistance films of Pt provide heat for the centrifuge cone and for a jet of air impinging on the surface of the iiquid being evaporated. Moisture and acid vapors from the chemical operations are pumped from the main box and adsorbed on the molecular sieves in a separate enclosure. The dry air is then returned to the main box. Thus, by capturing vapors within a closed system and by continuously recirculating the box air through the absorbers, a reasonably dry atmosphere is maintained at all times within the chemistry enclosure, and corrosive action is effectively reduced. (auth)

Two new explosives are under development at Lawrence Radiation Laboratory (LRL) which show considerable promise as replacements for PBX 9404 (and also PBX 9010). Three primary considerations have guided us in the development of these explosives. Compared to PBX 9404 we have sought a significant decrease it sensitivity, a substantial gain in mechanical properties and as little loss in energy and detonation velocity as possible. While testing is not complete on these materials and probably won`t be for quite some time sufficient information is available to warrant a complete status report at this time. An index of the topics covered herein is: Formulations and Raw Materials; Molding Powder; Explosive Performance; Sensitivity; Thermal Stability; Compatibility; Mechanical Properties; Machinability; Pressability; and Comparison of the Two Materials.

It is now generally known by those working on Tory II-A that the average maximum fuel element wall temperature obtained during the Intermediate Power Test was somewhat higher than the 2250 degrees F design value. The purpose of this report is to explain how this occurred.

Because of various evaluation studies on fuel stability, it is proposed that Coors Porcelain Company change their recipe for fueled tubes. Specifically, zirconyl nitrate should be used. in addition to uranyl nitrate. The detailed changes to be made are provided.

This article describes a test procedure for a liquid high explosive. The liquid to be tested is NTN, a mixture of nitromethane, tetranitomethane, and 1-nitropropane in 52, 33.3, and 14.7 weight per cent. The test is to study the sensitivity of the explosive to shock by firing it in projectiles from a M1A1 155mm howlitzer, with a charge such that it will not exceed 16,500 g`s acceleration.