Report documenting research and developments made by the Maritime Reactor Program of the Oak Ridge National Laboratory.

This technical report describes development work done on zonal centrifuge systems at the Oak Ridge National Laboratory and the Oak Ridge Gaseous Diffusion Plant during the period July 1 to December 31, 1962 under the Joint National Institute for Health-Atomic Energy Commission Zonal Centrifuge Development Program. A basic purpose of this project is to develop new methods for isolating virus particles associated with cancerous cells and tissues. Three classes of rotor systems capable of separating particles ranging in size from whole animal or plant cells to protein or nucleic acid molecules on the basis of either sedimentation rate or density alone have been developed. Experiments with phage particles indicate the feasibility of large-scale virus isolation by continuous-flow centrifugation, followed by isopycnic banding in cesium chloride and velocity sedimentation in sucrose - all steps being carried out sequentially in the same rotor. Zonal rotors using the reorienting gradient principle for molecular separations have been tested to 141,000 rpm (formula). Previous work on zonal centrifugation and future plans for this program are discussed.

Report that "discusses the magnetic vector and scalar potentials, magnetic field components and their derivatives, and flux linkage for single current systems, and the mutual inductance, forces and torques between two such systems, whose axes are coplanar but not necessarily coincident" (p. 1).

Startup accident and pump run-down on the ORR have been simulated on the Reactor Controls Analog Facility. At full flow the 150% level scram (45 Mv) easily terminates the startup accident before the metal temperature gets above 180°F. For very low flows typical of criticality runs, temperature coefficients turn the excursion before it reaches 150% of full power, and temperatures climb to boiling, a potentially hazardous condition. (This same behavior can occur at full flow is the power is increased to the point where the level scram must be set above 50 Mw).

A generalization of Wigner's simple model for the distribution of nuclear level spacings is studied. It is based on a stochastic process which reproduces the correct joing probability distribution of N enerfy levels for small spacings. The case N=3, which includes the effect of the correlation between adjacent spacings, is discussed in detail. The resulting distribution and the correlation coefficient are compared with experimental data. No definite conclusion can be drawn except to say that the effect of the correlations on the spacing distributions on the spacing distribution is very small.

This report is a working document of the UB AEC Nuclear Cross Sections Advisory Group and is issued about twice a year for the purpose of stimulating important cross section measurements. The requests listed in this report have originated from various Atomic Energy Commission contractors, the Advisory Committee on Reactor Physics, and the Nuclear Cross Sections Advisory Group. The distribution is limited to members of the US Nuclear Cross Sections Advisory Group (NCSAG), the US Advisory Committee on Reactor Physics (ACRP), the Tripartite Nuclear Cross Sections Committee (TNCC), the European-American Nuclear Data Committee (EANDC), to US AEC off-site contractors, and to the requestors and measurers of the cross sections listed in this report. The contents of this document are not to be republished in part or in full without special permission of the chairman of the NCSAG, The unpublished data which appear in this report must not be quoted in publications without permission of the experimenter.

An experiment was performed at the ORR in order to find a good fission chamber location. Two locations on the pool side of the reactor tank were explored with a one inch diameter fission chamber. The thermal neutron flux attention was found to vary nearly exponentially with distance, and no shadowing effect could be seen during a reactor startup. The fission products high gamma flux could be discriminated without difficulties. Both positions seem to be adequate to locate a reactor control fission channel.

The calculation of concentration transients caused by mixing in the interconnected three-body system of the fuel dump tanks, core and blanket has been coded for the IBM-704. The system of equations includes material balances, the dependence of critical temperature on core and blanket concentrations, and the effect of temperature (density) changes on flow rates. The controllable flow rates, the fuel dump tank weight and the initial conditions are input variables. Outputs include core and blanket concentrations, critical temperature, concentration ratio and power ratio. The code has been used to compute transients during a number of operating situations.

Experimental determination of heat-transfer coefficients, burnout heat fluxes, and friction factors have been made for swirl flow of low-and moderate-pressure water through electrically heated aluminum, nickel, and copper tubes containing full-length Inconel twisted tapes. For nonboiling conditions, swirl-flow heat-transfer coefficient were successfully correlated with both the Froude modulus (the ratio of inertial to centrifugal forces) and a grouping of the Grashof and Reynolds moduli (ratio of buoyant to inertial forces).

The addition of a surface active agent to an aqeous-organic interface produced a resistance to mass transfer equivalent to slightly more than 1 cm of water. Five semicontinuous Druhm runs were made with 1/2in. thick MgO liners and terminated due to either failures of the UFe nozzle or a top gasket leak. In preliminary scale-up tests of the flame calcination equipment, a maximum feed rate equivalent to 720 g oxides/hr was achieved using a 3-in. i.d. magnesia reflector with an outside wall temperature of 1500°C.

This memorandum summarizes the results of an analysis to determine the extent of displacement of the EOCR core due to blowdown in case of several postulated hot main gas coolant pipe failures. Results show that the core will be damaged for any hot pipe double-ended failure. Excepting the improbable case of no coolant flow existing proper to the break, the core will be damaged for any hot pipe fracture exposing a total flow area to the atmosphere equal to that of one pipe. Smaller breaks will probably be safe in this respect.

Report documenting the ongoing research and developments of the Oak Ridge National Laboratory's Homogeneous Reactor Project.

Report documenting the systems, operating procedures, and design of the Oak Ridge National Laboratory's gas-cooled reactor.

An estimate has been obtained of the heat that should be generated in the HRT core in order to hold the system at operating temperature under no-load conditions. This estimate was made by measuring the feed-water rate to the package boiler during an oxygenated water rung. Results are summarized.

The expected performance of the HRT carbon beds was calculated for various reactor operating conditions. the calculations indicated that the flow rate of sweep gas will have to be limited to prevent excessive activity discharge. Data on activity discharge are included.

During the pre-startup phase of the HRT operations, moisture was accidentally introduced into the charcoal bed adsorbers in the off-gas system. Tests have been made to determine the effect of wetting upon the adsorptive properties of the charcoal. The work was divided into two phases, testing of fresh charcoal in the laboratory and testing of the HRT charcoal beds in situ. It is recommended that the beds be dried by heating them to about 40 C and purging each with one to two liters/min of dry instrument air.

The measured pressure drop across the reactor core, with the sample holder in place, is 6.9 psi, more than twice the estimated value. Better estimates, based on more rigorous mathematical analysis, should be possible for future problems of this type. The 2% density difference which produced the relatively high velocity of approximately 1 fps, in this experiment, will result from a temperature difference of about 8 C. It is concluded that the bulk fluid temperature near the sample holder will be less than 8 C above the average temperature at the same elevation in the core.

In order to evaluate the behavior of an APPR type absorber rod, an irradiation test program has been established. Approximately 21 more samples are planned for testing under this request. The request proposes testing a full size APPR-1 type control rod in the MTR. The objective of the test is to better evaluate the neutron absorbing material proposed for the APPR-1 control rod.

The operation and design of a high pressure recombination loop for the recombination of H2, D2, and O2 produced by the radiolytic decomposition of water which is used a solvent for fuel in the homogeneous reactors are presented.

Progress report of the Oak Ridge National Laboratory Aircraft Nuclear Propulsion Project providing updates on various projects, experiments, and other work. This report includes summaries of project activities in: aircraft reactor test design, ART physics, ART instruments and controls, component development and testing, procurement and construction, ART, ETU, and in-pile loop operations, phase equilibrium studies, chemical reactions in molten salts, physical properties of molten materials, production of fuels, compatibility of materials at high temperatures, chemistry, analytical chemistry, metallurgy, dynamic corrosion studies, general corrosion studies, fabrication research, welding and brazing investigations, mechanical properties studies, ceramic research, nondestructive testing studies, heat transfer and physical properties, radiation damage, fuel recovery and reprocessing, critical experiments.

Progress report of the Oak Ridge National Laboratory Aircraft Nuclear Propulsion Project providing updates on various projects, experiments, and other work. This report includes summaries of project activities in: aircraft reactor test design, ART physics, ART instruments and controls, component development and testing, procurement and construction, ART, ETU, and in-pile loop operations, phase equilibrium studies, chemical reactions in molten salts, physical properties of molten materials, production of fuels, compatibility of materials at high temperatures, chemistry, analytical chemistry, metallurgy, dynamic corrosion studies, general corrosion studies, fabrication research, welding and brazing investigations, mechanical properties studies, ceramic research, nondestructive testing studies, heat transfer and physical properties, radiation damage, fuel recovery and reprocessing, critical experiments.

Report issued by the Oak Ridge National Laboratory discussing the development of a closed-loop control system. As stated in the statement, "the purpose of this thesis was to design and build a closed-loop control system to be used as an indicator of the level of high temperature liquids contained in a tank" (p. 1). This report includes tables, illustrations, and photographs.

A capital cost breakdown of the Idaho Chemical Processing Plant, a directly maintained remotely operated plant for processing spent enriched uranium fuel assemblies from reactors, is presented. The capital investment in the plant, including design, construction, training, and preoperational costs, an estimate of the direct costs incurred by the Atomic Energy Commission, and a proportional part of the costs of Central Facilities, including the value of the land and improvements theorem when acquired by the Commision, was $31,105,899. The cost of design and construction was $25,212,231, of which $3,773,357 was expanded on design and inspection.

Technical report outlining various experiments that took place in the Oak Ridge Area Critical Mass Laboratory until August 31, 1950. Experiments took place in categories of critical experiments, experimental engineering, heat transfer, radiation damage, nuclear measurements, reactor physics, ARE reactor design, circulating fuel reactors, and circulating moderator reactors. [From Summary]