Procedures for assembling EBR-II fuel elements with annular sodium bonds between the uranium rods and the stainless steel claddings are outlined. The results of several meltdown and uranium-settling experiments are given. Bonding experiments were performed: furnace bonding, submerged canning, ultrasonic bonding, centrifuging, pressure pulsing, and vibratory bonding. Vibratory bonding was chosen for the production of the first EBR-II core. (D.L.C.)

Spert IV is a large pool-type experimental facility for reactor kinetic studies. These studies will include power excursion and instability tests for a variety of reactor designs. Since the Spert IV experimental program requires the performance of tests which will approach, and may exceed the threshold of reactor destruction, the probability of occurrence of the maximum possible accident is not negligible compared with that of other possible accidents. The maximum possible accident for this facility is considered to be a severe nuclear excursion which results in the destruction of the reactor building and the release of 100% of the accumulated fission product inventory of the atmosphere in a steam cloud. The fission product source assumed in the analysis of this accident is an upper limit in view of the nature of the tests to be performed and the heat removal capacity of the system. This postulated accident is independent of the details of core and control system design and is valid for all cores anticipated for use in the experimental program. The major hazards present in the operation of this facility, the precautions to be taken to reduce the probability of an accident, and the consequences of the maximum possible …

A kinetic study was conducted to provide information on oxidation of reactor graphites in the temperature range of 450 to 675 deg C and on the effects of reactor environment on oxidation rates. Among the parameters studied were chemical reactivity of the graphite, prior oxidation, a high intensity gamma flux during oxidation, variation of the surface-to-volume ratio of the graphite specimens, neutron bombardment prior to oxidation exposure, and gas flow rates. Rate equations showed apparent activation energies of 50 kcal/mole in the absence of radiation and 30 kcal/mole in the presence of a 1 x 10/sup 6/ r/hr gamma flux. (auth)

Several diluents for possible use in TBP Purex Plant solvent were tested. The tests included nitric --nitrous acid degradation, fission prcduct distribution under simulated plant conditions, emulsillcation, and radiolysis. The order of quality of four diluents is n-dcdecane> Soltrol 170> Shell Code 85030(82000)> Shell E-2342. (D.L.C.)

The physics design problems of the EBR-II are summarized. These include analysis of the EBR-II engineering design as well as applicable zero-power critical experiments. Pertinent reactor safety problems are reviewed. Safety considerations bearing on normal plant operation and manipulations within the reactor are emphasized. The implication of controlled in-pile meltdown experiments is considered. Irradiation damage and metallurgical phase phenomena are summarized and related to reactivity. The nuclear performance of the system is considered in terms of actual plant operation. The predicted shift of both power and reactivity from core to radial reflector is described. (auth)

Integral physics data for fast reactor design are discussed. The measurements needed include those of critical mass, shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi- alpha , and similar quantities. Topics covered include Pu- and U/sup 233/-fueled systems, highly enriched U/sup 235/ systems in optimum geometry, uranium cores of various enrichments and dilutions, extreme geometry critical experiments, specific reactor systems, core mockup inhomogeneities, spectral studies and detector ratios, uranium equilibrium spectrum data, materialreplacement measurements, fast reactor dynamics, and suggested future experiments and experimental programs. (M.C.G.)

ABS>The design concept of an a-c ionization chamber and its supporting electronics is described. Several designs are possible and the sensors can be tailored to specific requirements when necessary. Mode of operation, signal voltage development, and switching frequency are discussed. High-sensitivity operation is described. Requirements for high-temperature, power-level operation are outlined. (M.C.G.)

Variations in the distribution of steam bubble, the "void" distribution, in a boiling channel as a function of changes in heating power were studied. A rectangular test tube, of 1.11 x 4.44-cm cross section and 127-cm height, was inserted in a forced-circulation pressure loop. The tube was heated by passing an a-c current through the tube walls. A power oscillator was built which could give a 10% peak-topeak sinusoidal power modulation at any frequency in the interval from 0.01 to 10 cps. Variations in the volume fraction of steam were observed by means of a gamma densitometer built for the purpose. Accurate void profiles could be taken by traversing the test channel vertically and horizontally. With the void detector stationary at a given height, the amplitude and phase delay of the steam void variations were measured in the frequency range mentioned. The signal from the gamma detector was passed to a harmonic analyzer built for the experiment. This instrument could pick out the void variations coherent with the power variation in the presence of much greater random signal variations caused by the boiling process. The frequency response of steam void was measured at 4 different pressures ranging from 27.2 to …

A bibliography on the electrical properties of glass is presented. The 267 references covering the period from 1930 through 1960 are arranged according to subject. An author index is included. (M.C.G.)

A study was made of the phase relations and the properties of uranium-- fissium alloys which have compositions bracketing that intpnded for the first core loading of Experimental Breeder Reactor II. The fissium aggregate in the alloys consisted of the elements Zr, Nb, Mo, Ru, Rh, and Pd. Phase relations are shown to parallel closely those in the dominant U--Mo--Ru ternary system. The uranium gamma phase is stabilized down to 552 deg C, while the beta phase is entirely suppressed at high fissium contents. Certain crystallographic data are given and the minor phases that occur in the alloys are identified. In cast and gammaquenched alloys the retention of the high-temperature gamma phase produced low hardness and low density. The thermal expsnsion behavior of the alloys is shown to be dependent upon composition and prior thermal history. Thermal conductivity data are presented for uranium and the uranium-- fission alloys. The thermal conductivities of the alloys decrease with increasing fissium concentration. (auth)

A procedure was developed for the preparation of uranium(IV) nitrate solutions in dilute nitric acid. Zinc metal was used as a reducing agent for uranium(VI) in dilute sulfuric acid. The uranium(IV) was precipitated as the hydrated oxide and dissolved in nitric acid. Uranium(IV) nitrate solutions were prepared at a maximum concentration of 100 g/l. The uranium(VI) content was less than 2% of the uranium(IV). (auth)

LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)

S>Criteria for stress analysis and structural design with steel for the critical components of nuclear plants are presented. An effort was made to integrate the effects on the strength of steel of the coexisting phenomena, such as mechanical and thermal loads, stress cycling and fatigue, creep and creep rupture, irradiation, and loss of ductility. Extensive use of the plastic region of steel was made for the accommodation of thermal stresses. The concept of cumulative damage in the plastic region was expounded for thermal fatigue and creep. A short description is given of the five avenues followed for the development of a theory governing the strength of materials. An approach was taken up that attempts to establish a "theory of fatigue" based on experiments. (auth)

Tabulations are included for: vibrational and rotational parameters for diatomic molecules; quadratic, cubic, and quartic vibrational force constants of diatomic molecules; parameters for empirical functibns relating force constants to bond length; and cubic force constants for bond stretching in polyatomic molecules. (B.O.G.)

Photograph used for a story in the Daily Oklahoman newspaper. Caption: "A Red, White and Blue birthday greeting is received by W. E. Harber, Shawnee banker, from his daughter, Mrs. David Mayne."

Photograph used for a story in the Daily Oklahoman newspaper. Caption: "Deserted by customers, a bored clerk in the jewelry department of Government Employes Exchange Sunday glumly pondered "essentials" and "non-essentials."

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A corrosion testing program designed to estimate the potential service life of aluminum alloys used in the construction of the Oak Ridge Research Reactor (ORR) cooling systems has been in progress for over two years. The five alloys (1100, 3003, 5052, 5154, and 6061) used to the greatest extent in the reactor exhibited continuously decreasing corrosion rates since the first 500-hr inspection. Samples exposed in the core-cooling loop have shown a decrease in corrosion rate from a 2.6 mpy maximum for one group during the first 500 hr to an over-all average of less than 0.1 mpy for another group after a full year in test, with the maximum metal loss less than 0.1 mils. Results indicate that with suitable water treatment the aluminum alloys used in the ORR may be expected to give satisfactory performance for many years. Based on the generalized corrosion rates alone, 40 to 50 years of service life may be expected. However, since occasional localized corrosion has been observed (rarely), minor repairs will almost certainly be required before that time. (auth)

Monthly newsletter from Dallas, Texas published by the Stamps Quartet Music Company that includes news and information concerning gospel music along with advertising.

List of engineers employed in the 11th District of the Missouri-Kansas-Texas Railroad South Texas Division as of July 1961. It includes the name of each person and employment date, with the most senior person listed first.

List of firemen employed in the 11th District of the Missouri-Kansas-Texas Railroad South Texas Division as of July 1961. It includes the name of each person and employment date, with the most senior person listed first, as well as seniority dates and new seniority dates for firemen barred from preferred service.

List of brakemen employed in the 7th District of the Missouri-Kansas-Texas Railroad South Texas Division as of July 1961. It includes the name of each person and employment date, with the most senior person listed first.

List of conductors employed in the 7th District of the Missouri-Kansas-Texas Railroad South Texas Division as of July 1961. It includes the name of each person and employment date, with the most senior person listed first.

The maximum operating characteristics that can be achieved with the use of UO/sub 2/ as a reactor fuel are being investigated. Fuel assemblies are being fabricated for measurement of fission gas pressure in UO/sub 2/-filled fuel rods. A pressurized-water loop in GETR is being designed. A heat transfer burnout study is in progress to establish the upper limit of heat flux for studies of central melting in a fuel rod and fuel-cladding interactions. (M.C.G.)