In this technical report, firstly, the transfer function of a non-regenerative, base-driven transistor circuit is derived by applying the linear equivalent circuit method. The results are experimentally verified. Secondly, the idea of the inverse-gain-bandwidth is introduced as the criterion of the dynamic design. Thirdly, the stability factor is explained. Finally, it is shown that a system constructed by various types of transistor switching circuits is reduced to a long train of unit chains formed by delay units and wave-shapers, and the maximum allowable number of delay units in a unit chain is discussed. The results given here are consistent with the design procedures for transistor switching circuits established in the Digital Computer Laboratory, i.e., the emitter-follower logical circuits associated with restorers and flip flops.

Most of the publications in the open literature available to us agree that sodium does not readily react with graphite, though potassium, rubidium, and cesium readily enter between the planes of the graphite lattice. In view of the conflicting evidence available, it seems inadvisable without very extensive further study to consider the use of graphite in direct contact with liquid sodium, except under very mild conditions of temperature and irradiation, and then only with very pure sodium and completely graphitized graphite.

A review of helium sources, availability, consumption, and future production capacity indicates that the helium consumed in the operation of 100 helium-cooled reactors for 300-Mwe power plants over a 20-year period will run no more than a few per cent of the helium consumed in the U. S. in the next 25 years and less than 1% of the helium that is to be stored during that period under the helium conservation act of 1960 (Public Law 86-777).

Analyses are generally made at the ORR with a view toward determining the surface temperatures of fuel plates.

A literature study has been made of rates for the reaction of oxygen with high-purity artificial graphite. Values from a number of sources have been expressed on a common basis which provides approximate correction for the retarding effects of oxygen diffusion in the graphite pores. The corrected rates can be correlated by the equation k = 7.24 x 10^9 exp (-22 100/T), where k has units of weight fraction oxidized per hour and T as in °K. Effects of oxygen concentration, solid and gas-phase contaminants, and radiation on the observed rates are discussed. Methods for estimating rates and spatial distribution of heat release during graphite oxidation are presented. These should be of value in analyzing the hazard of a graphite fire following a coolant system rupture in a gas-cooled, graphite-moderated reactor.

Calculations indicate that single-transfer line concatenated pulse column systems can be operated with static pressures that are not excessive if a sufficient number of vessels are employed in the system. The required number of vessels can be attained by using a series of short columns or by using holdup pots in conjunction with a limited number of columns.

Data on waste volumes and heat generation of several reactor fuels which may be reprocessed in the Power Reactor Fuel Reprocessing Pilot Plant at ORNL are tabulated.

The thickness of an oxide film on phosphor bronze helices was determined by first establishing the oxygen content of the helix "as received" and after cleansing with nitric acid. Based on the assumption that the difference between these two values was the oxygen in the film, and that the film consisted entirely of cupric oxide, the thickness of the film was calculated from the density of cupric oxide, weight of the film, and surface area of film. A value of 1080 A was calculated as the thickness by this method.

The 0.045m UO2SO4, 0.036m CuSO4, 0.025 m H2SO4 solution (HRT fuel composition) was chemically stable during 1, 866hr of operation at 280ºC and 1500 psi. The system was pressurized by boiling a 0.4 gpm stream of the fuel in a titanium heat exchanger at 313ºC.

Nickel is more susceptible to mass transfer in a 100hr non-isothermal dynamic corrosion system than is Inconel when exposed to a NaCl-MgCl2-UCl3 (50.0-33.3-16.0 mole %) bath at a hot zone temperature 1800 F. No nickel mass transfer was observed in a 500-hr test at 1350 F, but Inconel showed some attack under these conditions. Inconel mass transfer was observed in both tests, being more severe at the higher temperature. On the bases of these preliminary tests, it appears that nickel is a more satisfactory container than Inconel for the chloride bath at temperatures in the region of 1350 F. The chromium is more susceptible to selective leaching from Inconel at 1800 F during a short 100-hr test (0.26% Cr in bath) than in a 500-hr test (<0.001% Cr in bath) at a lower temperature (1350 F ).

The friction characteristics of a full-scale straight-tube model of the ART fuel-to-NaK heat exchanger were determined experimentally. The presence of the spacers resulted in a transition to semi-turbulent flow at a Reynolds modulus of 350. this semi-turbulent flow persisted up to a Reynolds modulus of 5,000. The circumferential spacers were found to contribute slightly more than the radial spacers to the pressure loss in the heat exchanger.

A simplified method was developed for calculating wall-scatter gamma radiation escaping through a shield opening. The method was applied to the HRT and the results showed that next to the line of sight contribution, scattering of the wall of the shield opening was the main contribution to the dose at the rear edge of the shield. Design charts were prepared that give the dose as a function of the gamma source location with the reactor cell.

This report supersedes ORNL CF-57-2-1, which was an extension and revision of ORNL CF-54-6-72. Details concerning a recommended system of flow-plan symbols and drawing are given. The system is designed to identify the function of all major instrument components and to show schematically the operation of the instrument relative to the particular process. The system is used for identification and designation. The system is a modification of the Instrument Society of American Recommended Practice (RP 5.1).

A program of bench scale leaching and ion exchange testing for the recovery of uranium from liquors produced by leaching Utex ore (Monticello Stockpile 28) was conducted at Monticello, Utah from July 1953 to April 1954.

The successful completion of a program of experiments, including the Aircraft Reactor Experiment (ARE), has demonstrated the high probability of producing militarily useful aircraft nuclear power plants employing reflector-moderated circulating-fuel reactors. Consequently, and accelerated program culminating in operation of the Aircraft Reactor Test (ART) is under way. In order to adhere to the compressed schedule of the accelerated program, it is essential that the Atomic Energy Commission approve the 7500 Area in Oak Ridge as the test site by February15, 1955. This report summarizes the hazards associated with operating the contained 60-Mv reactor of the ART at the proposed Oak Ridge test site.

This report summarizes the research and development work carried out during December, 1959, and January and February, 1960, for Budget Activity 2790 - Separations Development for Non-Production Reactors. The major effort on Activity 2790 has been completed. Current efforts on the remaining problem areas will enable Hanford to begin reprocessing fuel elements from power reactors which employ depleted or slightly enriched uranium fuels in July, 1962.

Ultrasonic cleaning uses high-frequency sound waves to induce cavitation within a cleaning, medium. During cavitation, millions of small bubbles form and collapse, resulting in agitation proportional to the energy put into the solution. The making and breaking of these bubbles produce the scrubbing action associated with ultrasonic cleaning.

One of the problems encountered in the design of the EGCR has been the possibility of bowing in the outer six rods of the seven rod cluster. The power distribution in an outer rod is asymmetrical with respect to the center of the rod. This causes a variation in the cladding temperature, and might result in unven thermal expansion, or bowing.

This memorandum clarifies the present status of ξ measurements in the uranium-water lattices.

This paper discusses a remotely operated system developed to transport (raise and lower) a 1/2 gram radium gamma radiation source for calibration of the HAPO film badges and finger rings. The system employs the rotometer principle for positioning the source for operating purposes. An accurate timer is utilized to assure the desired exposure time.

Presently rotary drum filters are being used for plutonium oxalate slurry filtration in the 234-5 Building. Changing of the filter cloth used on a rotary drum is a time consuming operation which involves ever increasing radiation exposure to maintenance personnel. Consequently, studies were conducted in the 321 Building on a disk type filter which could be adapted for simple and quick ("one-nut") replacement of the filter medium.

The Non-Production Fuel Reprocessing Program involves the chemical processing of valuable reactor fuels received from privately owned power reactors. It is necessary therefore, to accurately measure the fuel material received in order to insure proper payment to reactor operator and to provide the Atomic Energy Commission with appropriate accountability data. The volume measurement study described herein was conducted in order to determine the limits of accuracy that could be obtained in measuring relatively large volumes of solution under plant processing conditions.

The transfer or circulation of raw dissolver solutions containing gross particulate matter presents many problems not easily overcome by standard pumping equipment. In April of 1956 the HAPO concept of a modified Archimedes screw pump was developed. Two basic models, externally powered and driven through extended shafts, were constructed and tested. Operation of these preliminary models was so satisfactory that a third unit, integrally formed into drive motor, was built and placed in extended life test. This report describes construction and testing of the third and final model.

The final percentage of xenon created by fission in uranium and plutonium is a function of the neutron flux intensity. The flux dependence results because axenon 133 and 135 can be converted to a a stable xenon isotope by neutron capture instead of decaying into cesium.