The thermal-neutron fluxes in the lattice positions and selected experimental positions of the ETR were measured for Cycle 51B. Gamma heat generation was also measured in various experimental positions for Cycles 51B and 54. Vertical traverses for each position monitored are shown. (auth)

Various methods and techniques of measuring void fraction in boiling heat transfer media are discussed. Details of 17 void-fraction measurement systems are abstracted. A dynamic void-fraction measurement system is described, capable of continuously measuring void fraction at any position along a 12-ft test section. Void fraction (the ratio of vapor volume to total volume of a 2- phase mixture) is measured in boiling organic coolant contained by a test section of 5/8-in.-OD nickel tube with 0.049-in. wall thickness. Measurement of void fraction without physical contact with the test section is achieved through the use of a radiation attenuation method. The intensity of a narrow 100-kev x-ray beam transmitted through the test section is measured and recorded to permit calculation of void fraction. An x-ray tube and scintillation detectors are mounted on a remotely controlled elevator assembly used for horizontal and vertical positioning. Major sources of measurement error due to fluctuations in x-ray tube output and fluctuation of detector power supply voltage are cancelled by using two detectors connected in a differential circuit. Detector cooling and heavy magnetic shielding further reduce measurement errors. Typical measurement accuracies for small and large void fractions are 0.05 surface proces 0.015 and 1.00 surface proces …

Progress is reported for BORAX-V, liquid-metal-cooled reactors, nuclear physics, fuels development, chemical separations, plutonium recycle program, Advanced Research Reactor, magnetohydrodynamics, nuclear safety, and TREAT. (D.L.C.)

The theoretical development, statistical treatment, and application of hot channel factors in compact SNAP reactors as represented by the SNAP 8 experimental core loading are presented. The channel and rod power effects of variation in the concentrations of uranium, hydrogen, and poison in the fuel rods are given. The statistical distributions of all variables are examined. A random selection of rods and channels is compared to the actual SNAP 8 configuration, with no difference noted. Confidence limits are set on hot rod and hot channel factors. Various other factors and aspects of the problem are discussed. The results of the study showed that the hot channel factors may be reduced to less than +2% over the nominal power, as opposed to a hot channel factor of about +10% as previously determined by empirical methods. (auth)

Pressure drop and vibration characteristics were determined for an 8-rod fuel element model of the design intended for use with uranium carbide (UC) in the Hallam Nuclear Power Facility (HNPF). Measurements with water as the test fluid were converted to equivalent values for sodium, the HNPF coolant, using the principles of dimensional similitude. Initially UC elements will be concluded in an HNPF core loading comprised primarily of 19-rod U-Mo fuel elements. In this core loading, the UC fuel element requires 17.5 lb/sec of sodium coolant at a core pressure drop of 11 psi. The measured fuel element pressure drop ranged from 0.27 to 5.8 psi over the sodium flow range from 3.5 to 17.4 lb/sec. The existing HNPF variable orifice can adjust flow for this fuel element over the range from 5.7 to 21 lb/sec at a core pressure drop of 11 psi. No significant vibration of the fuel rods was induced by the flow of water. (auth)

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Information required to obtain authorization to receive and use a fast critical assembly for operation with the General Atomic linear accelerator is documented. The analysis offers evidence showing that operation of the assembly will create no undue risk to the health and safety of the public or the personnel of the Laboratory. Experiments with the accelerator pulsed fast assembly will consist primarily of measurements of neutron yield and pulse shape produced by the electron pulses from the Linac impinging on the spherical metal reactor. (N.W.R.)

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In studies of the effect of alloying elements on the rate of dissolution of Al in mercury-catalyzed nitric acid, it was observed that Co, Ni, and Si present at concentrations of 1 to 2% have significant passivating effects. Fe was slightly catalytic. Passivation due to Si was partly overcome by contacting the passive alloy with active, high-purity Al. Increased catalyst concentration improved the rates when Ni and Si were present. Neither approach was effective when Cu was the passivating element. For application to nuclear fuel reprocessing, it is suggested that for minimum reprocessing costs the potentially passive Al alloys not be used in fuel elements or that, failing this, electrochemical activation techniques be applied at the processing plant. (auth)

A test loop was designed and fabricated for the study of the heat transfer and fluid flow characteristics of boiler and superheater fuel elements at heat fluxes of 70,000 to 500,000 Btu/hr-ft/sup 2/. The design and operation of the loop are described. D-c resistance heating at low voltages was used. Control, instrumentation, and shakedown tests are discussed. Runs performed to date are described. (D.L.C.)

The objectives of the SNAP-7D program were to design, manufacture, test and deliver a thirty-watt electric generating system for a modified U. S. Navy NOMAD-class weather buoy to be stationed in the Gulf of Mexico. The sixty-watt Sr/sup 90/ thermoelectric generator, the relay panel, the batteries, and the installation of the system in a boattype buoy are described. In addition to delivering the power supply, many tests were required for the SNAP-7D system to demonstrate its conformance to the contract statement of work. The electrical tests of the generator and of the system, the shock and vibration tests, and the tests at the environmental temperature extremes are discussed. (auth) Quantitative studies were made of the angular distribution of elastic and inelastic scattering of 25-kev electrons on noble gases, and the results were compared with scattering theory in the Born approximation. Results are presented in tables and graphs. (M.C.G.)

Progress made in the development of the Sioux Falls Power Reactor is reported. Information is presented on stainless steel corrosion in oxygenated steam, development of low-enrichment superheater fuel elements, nuclear analysis of first-core boron-stainless steel poison shims and transient simulator, and testing and stability of oscillator rod assembly and in-core ion chambers. (N.W.R.)

Research and development progress is reported on water-cooled reactors, liquid-metal-cooled reactors, general reactor technology, plutonium-recycle reactor program, advanced systems research and development, and nuclear safety. (M.C.G.)

Development progress is reported on fast reactors, liquid metal cooled reactors, ZPR-III, ZPR-VI, ZPR-VII, Faret, EBR-I, EBR-II, Borax-V, Treat, Fermi Fast Breeder Reactor, magnetohydrodynamic cells, thermionic cells, rocket fuel test reactor, and HiC critical assembly. Reactor technology is reported on fuel development and properties, heat engineering, physics, materials development, chemical separations, nuclear safety, and nondestructive testing. (N.W.R.)

Research and development activities are reported on Borax V, ZPR- III, - VI, and -IX, EBR-I, EBR- II, and FARET. Reactor technology investigations are reported in reactor physics, reactor fuel development, viewing systems, heat engineering, and chemical separations. Research on advanced systems is reported on reactors and undersea nuclear power applications. Nuclear safety development in thermal and fast reactors is also reported. (J.R.D.)

Fabrication of the components for the FDM-1 converter and its back-up device was completed. The FDM-1 converter was assembled, and preparations for testing were completed. Performance testing was initiated, and a total of 330 hours of operation above 1000 deg K in a cesium environment were logged. The peak emitter temperature reached was 2000 deg K with a thermionic output of nearly 2 watts/cm/sup 2/. Indications of a cesium leak were observed, and the device was shut down for examination. A 500-hour thermal endurance test of the clad fuel pellets was performed. A post-test examination was conducted, and no diffusion of UO/sub 2/ into the cladding was observed. A thermal conductivity test specimen was exposed to a temperature of 1550 deg C for 500 hours with the result that some intergranular diffusion was noted. The post-test examination of the vacuum emission test pellet was also completed. The design of the advanced double-diode thermionic converter utilizing the results of prior efforts was initiated. (auth)

References to 529 articles on nuclear reactor control, kinetics, and stability published before autumn 1962 are included. Emphasis is on calculations and theory since the references serve as an aid in analyzing the dynamic behavior of SNAP reactor systems. (D.C.W.)

Introduction: The effects of temperature on the mechanical properties of Zircaloy-2 containing up to 7.5 wt % niobium and 3500-ppm oxygen were studied and are presented herein.

This is the monthly report for the Hanford Laboratories Operation, July 1963. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This document discusses the leak repair of the 107-D retention basin. Leakage of effluent water from the 107-D Basin had increased from 6000 gallons per minute in November of 1959 to the peak of 8000 gallons per minute in April of 1963. Washout of fines had apparently created large voids beneath the north and west walls and endangered the integrity of the structure by settling the wall 2 in. at one place. The leakage was arrested in March of 1963 by filling the voids with concrete and grout and sealing the repairs with a newly developed sealant. Present total leakage is 3750 gallons per minute.

This is the monthly report for the Hanford Laboratories Operation, October 1963. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

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This is the monthly report for the Hanford Laboratories Operation, December 1962. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This document details activities of the irradiation processing department during the month of January, 1963. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; and Financial Operation.