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)

Hydraulic tests were made on the Hallam Prototype Free-Surface Sodium Pump to determine the net positive suction head (NPSH) required at various sodium flow rates. Pump performance data were also collected. The results indicate that an NPSH of 22 ft sodium is required at the design flow rate of 7200 gpm at approximates 1000 deg F, agreeing with computed values, and that the pump is designed with a safety margin of slightly over l0%. (D.L.C.)

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.)

A remote dissolution and analytical program for irradiated thorium is given. The aluminum jacket on the slug was dissolved with 6M nitric acid and 0.005M mercuric nitrate. After a water wash, the thorium dissolution was accomplished with concentrated nitric acid made 0.04M in hydrofluoric acid. Weighing, dissolving, and sampling were done remotely in the multicurie cell at the Idaho Chemical Processing Plant. Handling techniques for weighing and dissolving the slugs are described. Transferring and sampling apparatus as well as sampling techniques for the dissolved material are discussed. Analytical data obtained are tabulated. Abstracts of analytical methods for uranium concentration and isotope ratio, aluminum, thorium, cesium, and cerium are given. (auth)

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 …

Correlation of a number of physical and chemical properties of 12 different UO/sub 2/ powders was continued. The UO/sub 2/ powders were studied by means of infrared absorption measurements, oxidation temperatures as determined by hot stage microscopy techniques, and B. E. T. surface area measurements. Additional pellets were prepared to study ceramic performance. (M.C.G.)

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)

The types of tests performed in SPERT-I, and the tests proposed for SPERT-II and -III, are described. These reactors are described, and factors influencing their dynamic behavior are discussed. The tests are classed as static, step, ramp, and oscillatory. The correlation between the test results and the reactor dynamic safety characteristics (stability, self-shutdown under excursion conditions, etc.) is investigated. (T.F.H.)

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)

An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)

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)

The study of the formation mechanism of free radicals in polymeric materials was continued. Emphasis was placed on an examination of the effect of structural factors on the efficiency of free-radical site formation in acrylate polymers. Site measurements as a function of dose were made for polymethacrylamide and repeated for polymethacrylic acid. The volatile products from the irradiation of polyacrylic acid. polymethacrylic acid, poly-ter- butylmethacrylate. and polycyclohexyl methacrylate were measured quantitatively by mass spectrometry and vapor-phase chromatography. Grafting studies were initiated using polymethylmethacrylate as base polymer and vinylpyrrolidone as graft monomer. (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.)

The rapidly increasing volume ot new data in scientific and technical fields demands faster and better ways to communicate the new information to those concerned. A working system. System of Automatic Processing and Indexing of Reports (SAPIR), for doing this is described. SAPIR makes use of the Keyword-in- Context Index principle, by which certain keywords, together with surrounding words that act as modifiers, are selected from the titles of the technical publications. These keywords with their modifiers serve as index entries from each title, therefore there will be as many index entries generated as there are keywords contained in the title. The index entries are sorted alphabetically by the keywords. The first letters of the keywords in each of the index entries form a column in the printed format, which makes scanning for the soughtafter item easier. The SAPIR system is automatic and the list of citations is machine- generated on an IBM 1401 Computer. (auth)

The alloying of zirconium to improve corrosion resistance has an empirical basis, and satisfactory explanations for the alloying effects are not available. A theory of compensating valencies in the corrosion oxide is proposed, in which cations of lower and higher valence than zirconium (+4) are present in ratios such that electrostatic neutrality is ensured. An example is an alloy containing equimolar amounts of scandium (+3) and niobium (+5). A number of zirconium alloys were prepared in which scandium or yttrium were paired with elements capable of a +5 or +6 valence. The ternary alloys containing scanadium were superior to the alloys combining yttrium. The alloys containing scandium plus molybdenum, tantalum, or tungsten had relatively long lifetimes in steam at 540 deg C and 600 psi as compared with other alloy combinations, including Zircaloy-2. A quenched alloy containing 0.025 wt% Sc and 0.053 at.% Mlo, that is, 0.05 mol.% of each additive, corroded approximately according to a cubic law up to 758 hr, at which potnt the rate suddenly increased in a manner suggesting hydrogen damage. Examination of the oxide film from alloys containing scandium and molybdenum showed only monoclinic ZrO/sub 2/. It is believed that stabilization of this form of …

Experimental results were obtained on the effect of an ultrasonic field on the burnout heat flux for water flowing at atmospheric pressure, through an annular flow channel formed by a 1/4-in.-diameter electrically heated tube and a concentric glass tube of 3/4-in. ID. The active length of the central heating element was 5 1/2 in. The ultrasonic transducer, which was operated at 25,000 cps and a maximum electrical input of 300 watts, was located at the inlet end of the flow channel. The ultrasonic waves were propagated in the water in the direction of flow and thus parallel to the surface of the heating element. Burnout conditions covered channel inlet flows from 1.61 to 6.25 ft/sec and subcooling from 16 to 28 deg F. No effect of the ultrasonic field on the burnout heat flux or on the visible boiling phenomena at burnout conditions was detectable. During boiling at heat fluxes well below burnout, the effect of the ultrasonic field was a reduction in the diameter of the envelope of bubble activity surrounding the heating element. Visual inspectibn appeared to show that this reduction was associated with a smaller average bubble size and a greater frequency of bubble formation. However, all …

Results are presented of short time, high heat flux capsule irradiations to determine the magnitude of plastic strain induced in the capsule cladding due to UO/sub 2/ thermal expansion. Annular geometry capsules were employed with both internal and external cladding and forced convection cooling of both surfaces to simulate the reference superheat fuel element design. Both 304 stainless steel and 3003 aluminum claddings of various thickness were used to determine the influence of cladding strength. The initial gap spacing was purposely varied as an additional parameter. The program was undertaken as a part of the effort to evaluate plastic strain cycling as life limitation in the use of thin fuel element cladding. An over-all summary of the experimental results is included. (auth)