Annual report submitted by the Girl Scouts of the United States of America to Congress describing highlights from 1962, activities, organizational leadership, and other information about scouting programs.

The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.

The KER loops are four pressurized water-cooled in-reactor testing loops in which fuel, coolant and corrosion tests are being conducted for the N Reactor program. Two of the loops, KER-3 and KER-4, are equipped with N-size Zr-2 process tubes and have sufficient pumping and heat dissipation capacity to permit testing of 14-element charges of N fuel. The other two loops, HER-1 and KER-2, are fitted with Zr-2 process tubes of 2.1 in. inner diameter (vs 2.7 in. for N). The small loops can be used to test charges of N inner tubes or other elements sized at about 1.8 in. outer diameter or less. An inert sleeve must be used with N inner tubes to assure adequate coolant flow over both surfaces of the elements. The KE reactor active zone is 28 in. long com pared to 35 in. for N so the length of test charges is somewhat shorter than actual N charges will be. The purpose of this report is to review the testing environment, to discuss the testing program and to present the schedule as presently seen.

It is the purpose of this document to provide the scope of piping changes that are needed to supply gravity-fed water to the reactor storage basins on an unattended basis following complete reactor discharge of metal into these basins.

The objectives of this production test are: 1. To authorize the outage time involved in installing and removing oxidation monitors for the reactors, and 2. To specifically define sample design, spacing of samples in the channels, and residence times to insure adequate routine monitoring of moderator oxidation rates. This production test is concerned with B, C, D, F, DR, F, and H reactors. Similar requirements have been specified in PITA IP-20-I for KE and KW reactors

Hydraulic and deflection analyses of the unequal flow passages in the HFIR control region revealed excessive pressure differentials. A proposal to drill connecting holes between the flow passages initiated an experimental program for investigating the effectiveness crossflow on pressure equalization. Empirical constants for an analytical model and experimental design data on reduction of pressure differentials between channels were obtained. It was determined that crossflow between channels would satisfactorily reduce the pressure differentials. (auth)

The pre-Buggy experiment was designed to measure the fraction of vented radioactivity from a series of HE underground detonations containing radioactive sources. Capsules containing curie amounts of La/sup 140/ were sufficient to provide a radiotracing of the debris which resulted from the detonation. Particle size measurements of the debris indicated that La/sup 140/ was adsorbed on the surface of the soil particles. Some 96% of the activity was associated with sub-sieve particles representing only 8% of the mass and 90% of the available surface area. (auth)

Metals ---technology of high-temperature liquid, review; Systems for Nuclear Auxiliary Power (SNAP-2) ---coolant boiling research for; Nuclear Power Plants ---development research on thermionic, Los Alamos Molten Plutonium Reactor Experiments ---fuel element analysis; Heat Transfer Systems ---conference on liquid metal; Bearings --coating for liquid metal, performance of; Systems for Nuclear Auxiliary Power (SNAP-8) ---component development; Sodium---heat transfer research on; Mercury --heat transfer research on; Potassium ---heat transfer research on; Nitrogen Systems ---Hg --N, flow characteristics for two-phase; Mercury Systems ---Hg --N, flow characteristics for two-phase; Nuclear Power Plants ---heat transfer rejection by space, research on; Alkali Metals ---heat transfer research on; Turbines ---development of twostage potassium; Bearings --- development of liquid-metal lubricated; Niobium Alloys ---properties for alkali metal containment, research on; Rubidium ---thermophysical properties of; Los Alamos Molten Plutonium Reactor Experiments ---design of LAMPRE-1. (D.C.W.)

The corrections required to deduce unperturbed neutron fluxes from measurements with foil absorbers were determined experimentally for the case of gold foils exposed in a light-water medium. The results are in good agreement with the theoretical predictions of Dalton and Osborn and of Ritchie and Eldridge. (auth)

Activities of the radiation safety program at Oak Ridge National Laboratory for the period 1960-1962 are described. A reduction in the frequency of radiation accidents was observed and a training program for technical and nontechnical personnel has been formulated. Sixteen facilities were improved to ensure safe operation at a cost of approximately 388,589. A comprehensive Radiation Warning and Communication System is being installed in three phases over a period of three years ending in FY 1964. (auth)

The program, TENSOR, for the calculation of underground explosion phenomena and other time dependent problems involving the motion of elastic, plastic, fractured, and fluid materials is discussed. The method used in the code allows treatment of transient phenoraena in two space dimensions (cartesian or axial symmetry), and is particularly appropriate for problems involving compressible flow, large displacements, and free transitions of material between the elastic, cracked, plastic, and fluid states. The equations of motion are a straightforward generalization (to include a stress tensor) of conventional Lagrangian techniques for compressible flow and shock hydrodynamics problems. Methods for defining the stress system are discussed, and examples are included. (auth)

Design and operation of an electronic ripple filter to be used with the Brooknaven alternating-gradient synchrotron during flattop at energies between 10 and 30 Bev are discussed. (D.C.W.)

The effect of irradiation on the hydrolytic behavior of uranium monocarbide as related to the aqueous chemical processing of these reactor fuels is being investigated. The fabrication, evaluation, and encapsulation of the UC in niobium prior to irradiation is described. Uranium monocarbide buttons were prepared by arc-melting 1.51%enriched uranium metal with the stoichiometric quantity of spectroscopic-grade carbon. The buttons were then remelted and drop- cast into graphite thimbles 3/8 in. dia, and 3 in. long. The cylindrical castings were surface ground and cut into 0.17 in.-thick pellets, which were then polished. Preirradiation evaluation included metallography, chemical analysis, and hydrolysis studies. Microstructures of the final pellets indicated that the monocarbide was nearly stoichiometric; however, some alpha -uranium was present along with carbon contamination from the graphite thimbles. The use of tungsten- tipped electrodes in the arc-melting operation introduced about 0.3% of tungsten impurity. The composition of the pellets, as estimated from hydrolysis studies and chemical analyses was 92% UC, 3.5% UC/sub 2/, 4.0% uranium metal, and 0.3% WC. The UC pellets were encapsulated in niobium and subsequently shrunk-fit into a stainless steel cladding. The capsules were sealed by heliarc welding in a helium atmosphere chamber. (auth)

An experimental flight test was made to verify and improve the methods and techniques used to predict the aerothermodynamic effects on re-entering SNAP devices. The amount of aerodynamic heating as a function of body shape, size, and mode of entry was indicated. Test specimens yielded trajectory, aerodynamic heating data, heat distribution, and heat transfer of known bodies. IBM 7090 digital programs were utilized and so constructed that one digital code complemented the other, i.e., at any time during the re-entry period, the complete thermal history of the body is known. The test specimens consisted of lead bodies containing alkali metals, used as flaring material, and a larger titanium cylinder containing telemetry equipment to monitor the exterior and interior wall temperatures during re-entry. By judicious placement of transducer sensors, aerodynamic heating rates and mode of entry, i.e., tumbling, spinning, and/or stabilized, were determined. The lead bodies were consumed at high altitudes whereas the telemetry capsule survived re-entry. (auth)

Progress made during the period is reported including analyses of experiments and tests, and evaluation of effort. The subject material concerns reactor transient and excursion tests, fission product release studies, end-of- life shutdown devices, critical configuration studies, mechanical and thermochemical effects, reactor separation and fuel element ejection, and reentry burnup of fuel elements. The design of the SNAPTRAN 2/10A-1 and -2 machines was essentially completed and fabrication of major components was initiated. The fission product release tests at the NRTS were completed. Some progress was made on the explosive shutdown projectile design. A practical shipping sleeve design was demonstrated which will maintain the SNAP 2/10A assembly in a subcritical condition as a result of postulated shipping accidents. The Phase I ground test series were completed. The computer code, RESTORE, was shown to be excellent for the calculation of reentry trajectories. The calculation of fuel rod heating and ablation using a digital thermal analyzer program resulted in an improved and more elaborate analytical model. (auth)

Thermodynamic considerations indicate that sodium should be favorable for the one-step reduction of UF/sub 6/ to uranium metal. A reaction vessel was developed for the continuous reduction of UF/sub 6/ to metal with batch collection of the products, and several experimental tests established the chemical feasibility of this direct and continuous reduction. Up to 93.5% of the uranium content of UF/sub 6/ continuously reduced by sodium in a reaction vessel was recovered as massive uranium metal in the form of a metal button of acceptable purity. A remaining problem is development of compatible materials of construction. (auth)

The Fuel Cycle Facility for the Second Experimental Breeder Reactor (EBR- II), the process equipment, and the operations to be conducted in the facility are described. The Fuel Cycle Facility is a plant for reprocessing, by pyrometallurgical methods, the core and blanket material discharged from EBR-II. The reactor core alloy is uranium-5% fissium and contains about 46 wt.% U/sup 235/ . The blanket material consists of U in which plutonium is bred. Core and blanket subassemblies contained in transfer coffins are transferred between EBR- II and the Fuel Cycle Facility, which is in an adjacent building. The Fuel Cycle Facility consists primarily of an argon-atmosphere cell where fuel processing is done, an adjacent air-atmosphere cell where reactor subassemblies are assembled and disassembled, and an operating area (for personnel) which surrounds the two cells. Because of the high levels of activity expected, the fuel-handling-and- processing equipment is designed for remote operation. Remote processing is accomplished with the aid of bridge cranes, electromagnetic bridge manipulators, and master-slave manipulators. Transfer ports and air locks are used in the transfer of materials and equipment into the air-atmosphere cell and between the two cells. The walls between the argon-atmosphere and air-atmosphere cells and the operating …

Even though VBWR shutdowns were required for location and removal of five failed fuel assemblies (HPD Program), the increase in fuel exposure was good. Fuel exposures wili pass the values at which cold worked stainless steel cladding was failing under the HPD Program. Failure of the 0.005-inch cold worked stainless steel clad fuel rods in assembly 8L was traced to strain cycling fatigue. A study of tapered fuel rods indicates a potential advantage for us of a variable water/fuel ratio along the flow channel. Natural circulation tests in the hydraulic stability loop were conducted over a range of conditions from stable, to oscillatory with exponential decay, to self-sustaining oscillation of constant amplitude, to unstable oscillations with divergent amplitude. The response to impulses in power input shows the effect of the time delay for transporting steam voids up through the riser. The data permit calculation of oscillation frequency, damping coefficient, time lags, and show the magnitude and character of pressure and velocity changes. The data, which have an experimental scatter of plus or minus 10% maximum, show that burnout heat fiux: decreases with increasing flow up to 2 x 10/sup 8/ lb/hr-ft/sup 2/; has a maximum for hydraulic diameter between 0.25 …

A mathematical model for an associative memory is proposed that uses associative addressing and distributed storage. Associative addressing is accomplished by mapping from a space with relatively few dimensions (input variables) to the vertices of a binary-valued hypercube embedded in a much higher dimensional space. The dimension of the image space is chosen to be sufficiently grent that a hyperplane can be passed through the origin such that the relative distances to the image points are the relative functional values that are to be stored. The distributed memory is achieved in the n-tuple representation of the hyperplane, since each element will in general be used in calculating the distance to many points (images), and hence in storing many functional values. A technique formulated for solving the large linear systems that arise in such a problem and a proof of the convergence of such a procedure are included. Unfortunately, the basic form of an associative memory imposes the restriction that only a single linear expression be available at any one time, and that further its relation to other expressions not be known. This generally imposes a iurther restriction that the linear expressions be randomly drawn from the linear system and returned. …

Some simple models of (renormalizable) meson-nucleon interaction are examined in which the nucleon mass is entirely due to interaction and the chiral ( gamma /sub 5/) symmetry is broken'' to become a hidden symmetry. It is found that such a scheme is possible provided that a vector meson is introduced as an elementary field. (auth)

>The Piqua-Elk River spent fuel shipping cask, designed to transport 19 fuel assemblies from either reactor starting 120 days after reactor shutdown for refueling, is described. The maximum weight of the ioaded cask is 30 tons and the dimensions are 55 in. dia x 108 in. high. The space envelope required for the cask and shipping structure is 7 x 10 x 10 ft. The cask system is composed of 6 major components: cask body, closure head, fuel basket, shipping structure, lifting yoke, and Elk River fuel assembly adapters. Internal surfaces are stainless steel and external surfaces are painted carbon steel. The annulus between the inner and outer shells is filled with lead shielding. The fuel basket is an aluminum cylinder in which 19 cavities were machined to receive the fuel assembiies. The cask is designed for dry shipment at atmospheric pressure. (M.C.G.)

A survey of the,N. S. Savannah reactor shield was carried out during the report period. The results showed that the shield effectively meets design specifications. Tests of 12 removal by a charcoal unit in the N. S. Savannah reactor emergency ventilation system using 80 to 90% water-saturated air showed a unit efficiency of 99.86 plus or minus 0.07% at the 95% confidence level. Simplified procedures were devised for using shipboard equipment to screen personnel suspected of having body burdens of radioactive materials. Other work in connection with N. S. Savannah is reported on accidental activity release, port survey visits, and physics calculations. Continued trouble-free operation of the ORR pressurized-water loop is reported. Emphasis during the period was placed on water-borne radioactive contaminants. Advanced core developmental work was devoted to physics calculations, fuel development, fuel irradiation and postirradiation examinations, and evaluation of nonsintered UO/sub 2/ fuel for N. S. Savannah. (J.R.D.)

Practical procedures were demonstrated for partitioning plutonium from uranium in the Purex process with hydrazine-stabilized uranium(IV) nitrate as the reducing reagent for plutonium. Typical partitioning results were 0.04--0.2% loss of plutonium to the uranium product and less than 1 wt% uranium in the plutonium product. With uranium(IV) nitrate as a replacement for ferrous sulfamate, most solids in wastes from the Parex process can be eliminated. No undue hazard is expected from the use of hydrazine as the stabilizing agent. (auth)

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