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.)

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)

In the Plant Improvement Program, it is proposed to retube the K-Reactors with standard size Zircaloy-2 smooth-bore process tubes and to charge self supported fuel elements starting March 15, 1964. The first step in support of this transition program is to confirm compatibility of the fuel-tube geometry and secondly to obtain fuel-tube performance information prior to full scale commitment of the K-Reactors to this design. In view of the testing of self-supported fuel which has been accomplished to date and that which is planned, there is little incentive to install more tubes in a K-Reactor than are required to make the fuel-tube geometry check. To accomplish this, ten tubes are viewed as the maximum number that would be required. This report presents the design of a test to fabricate and irradiate ``K`` self-supported fuel elements in limited quantities.

Fuel element misalignment is an apparent cause of ``hot-spot`` ruptures. Several methods of eliminating hot spots have been tried, however, none appear to have completely solved the hot-spot problem. Attachment of projections to the side of the fuel elements appears to offer a means of minimizing misalignment since they act as bumpers against the side of the process tubes. Preliminary data from tests now in progress or recently completed indicate excessive corrosion rates are not to be expected on fuel elements with projection attached. In fact, no hot spots were found on 39 columns of normal self-supported I&E fuel elements discharged at normal goal or on four columns of enriched self-supported I&E elements irradiated to 850 mwd/t exposure.

Design studies were performed and costs were estimated for an air transportable, 10 Mw(t), pressurized light water thermal circulation reactor, combined with a direct conversion thermoelectric generator and static electrical inversion equipment. This TCR-TE'' concept appears to have potential for ultimate use as a remote unmanned power station. Based on an extrapolation of present reactor technology and on assumed thermoelectric materials properties forecasted to January 1, 1963, a net a-c electrical output of 315 Kw is estimated, assuming the use of 80 deg F local water for cooling purposes. An alternate concept using 80 deg F air for cooling produces 271 Kw, net. These electrical output figures can be improved significantly through a recommended research and development effort. The design and construction of a prototype plant is also recommended. (auth)

The lifetime of a B/sub 4/C-in-tubes'' control rod may be limited by either nuclear worth depreciation or internal gas pressure buildup as helium is formed and released from the B/sub 4/C particles. Nuclear life is enhanced by thin tube walls for any given O.D. of tubing and by high B/sub 4/C density, but the time to failure from gas pressure buildup is increased by the opposite. The optimum B/sub 4/C density and tube wall thickness were calculated for maximum control rod life, using several release fractions of generated helium and a number of allowable cladding stress values. It was estimated that lifetimes of l0 or more years for B/sub 4/C-in-tubes control systems are achievable through the proper selection of tubing material and B/sub 4/C density values. The B/sub 4/C densities required for maximum lifetimes are in a range that can be obtained by vibratory compaction followed by swaging. (auth)

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.)

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.)

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.)

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)

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 …

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.)

Technical report. From Abstract : "Several hundred long, thin-walled uranium tubes were extruded to close dimensional tolerance by a potentially low cost fabrication process in commercial facilities. The principal process steps were; forging uranium ingots, machining the forged ingots into billets, and extruding the billets into tubes. Some 750 tubes (maximum dimensions: 3.5 inches in diameter and 15 feet long) were fabricated by this process. Dimensional control was excellent in that 88% of the tubes were within +/- 0.010 inch of average dimensions over their entire length."

Abstract: This report describes the fabrication of elements for the OMRE third core loading.

From introduction: "The purpose of this report is to present the details of the exploratory and developmental work on tandem rolling, and the subsequent fabrication of tandem rolled fuel rods for irradiation testing in the Vallecitos Boiling Water Reactor."

The raw water export system forms the last ditch water supply system to the ``O`` and ``C`` type reactors; in the event of electrical and steam power failure, the export system is designed to supply enough raw water coolant. After the original export orifice was modified twice, the export system was retested.

This report details activities of the Fuels Development Operation for the months of April, May, and June 1961.

Production output of fuel elements during June was 108% of forecast. A new record high yield of 91.2% for 6-inch elements was established. Engineering of present reactor fuels and N-Reactor fuel development are described. (DLC)

The ACIR Library is composed of publications that study the interactions between different levels of government. This document addresses governmental structure, organization and planning in metropolitan areas.

The ACIR Library is composed of publications that study the interactions between different levels of government. This document addresses governmental structure, organization and planning in metropolitan areas.

The monthly report for the Hanford Laboratories Operation, June 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, operations research and synthesis operation, programming, laboratory auxiliaries operation, and professional placement and relations practices are discussed.

Report that summarizes multiple laboratories' reports on global fallout deposition. Reports include data on Strontium-90 deposition recorded by the Health and Safety Laboratory, data from other laboratories, related interpretive reports, and recent publications related to fallout.

Abstract: "Weights of linear high-explosive charges fired on the Yucca Lake playa of the Nevada Test Site varied from 0.23 to 42.7 pounds per foot. Crater and ditch dimensions and volumes resulting from these shots, fired during the fall of 1959 and spring of 1960, are presented here as a function of charge burst depths. Scaling relationships determined are as expected; i.e., square-root scaling of linear-charge weight per foot for ditch width and depth and a direct linear-charge weight relationship to ditch volume are obtained. Permanent ground surface displacement varies as the -3.22 power of the scaled distance from the charge. Detonation effects, charge shape effects, ditch erosion, and ditch cross sections are discussed. The appendixes present Toboggan data, results of soil investigation, and mathematical treatment given the data."

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.)