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.

EXPIRE is a calculation which predicts the reactivity-lifetime, instantaneous and integrated effective multiplication constants and instantaneous and integrated effective multiplication constants and instantaneous conversion ratio for heterogeneous reactors. The concentration of all the isotopes of interest from Th232 to Am243 are calculated as a function of time using the average reactor power density and a uniform flux distribution. The equations have been programmed for the IBM-704 computer and the average running time is approximately two minutes per reactor lifetime.

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).

Report documenting ongoing research and developments at the Oak Ridge National Laboratory's Gas-Cooled Reactor Project.

The hardness of several materials that have been considered for use as valve scats and bearings in thorium oxide slurry systems were measured and are reported for comparison with thorium oxide.

A helium coolant purification system was designed for the proposed 800 MWT Pebble Bed Reactor. The purification system will operate on a coolant side stream with a flow rate 1% of the total coolant flow and there are provisions for radioactive and non-radioactive contamination removal.

A helium coolant side-stream purification system consisting of parallel sections for radioactive and non-radioactive de-contamination was designed for the proposed Pebble Bed Reactor Experiment. Primary equipment components are two gas coolers, gas heater, charcoal delay trap, CuO oxidizer, Molecular Sieve adsorber, and full flow filter. The charcoal delay trap is sized to provide a hold-up of 30 minutes for Kr isotopes, 6hr hold-up for Xe isotopes, and 99.9% retention of iodine isotopes resulting in "de-contamination factors" varying from l for Kr85 to 556 for I131. Non-radioactive de-contamination will result in a steady state concentration of CO2 in the coolant of 20.8ppm or less.

The attached memo by G. N. Krouse of Sturm-Krouse, Inc. gives results of a preliminary analysis of the deflections of beam holes due to thermal expansion and internal pressure in the vessel. A partial solution of the problem is suggested. Based on preliminary pressure-temperature data the following deflections were derived: Movement of horizontal beam tubes = 0.046 in. Movement of Engineering facility tubes = 0.117 in. Vertical motion of the vessel at the horizontal beam tubes due to thermal expansion may be eliminated by locating the supports in that plane. That also will reduce the expansion at the point where the slant tubes pierce the vessel wall.

Current status of the High Flux Isotope Reactor which is being planned for construction at Oak Ridge.

Multigroup one-dimensional calculations were done recently to obtain estimates of critical masses, power density distributions and fissioning spectra for some homogeneous molten salt reactors having outer reflectors and central "islands," placed inside the currently proposed MSRE vessel. For a 5-inch-thick outer reflector and 1-ft-diamter island, both beryllium, the calculated critical mass is 108 kg; 40 percent of the fissions occur at thermal, and the maximum power density of 3.9 times the core mean power density occurs at the island-salt interface. If the reflector thickness is increased to 10 inches, the critical mass is reduced to 34 kg; 67 percent of the fissions occur at thermal, and the peak power density of twice the core mean again occurs at the core island-salt interface.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Program. Descriptions of progress and studies conducted are presented. This report includes tables, illustrations, and photographs.

From Summary: "The major objective of the run was the investigation of fuel stability. The reactor operated for long periods at the design power of 5 Mw with none of the usual indications of instability while the system pressure was kept at 1250 psig. Run 21 ended on January 22, so that the reason for an abrupt change in the mixing rate of fuel between the core and blanket could be investigated and a reactor steam-system valve could be repaired."

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Program. It includes descriptions of progress and studies conducted during the report period, with tables, illustrations, and photographs.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Program. Descriptions of progress and studies conducted are presented. This report includes tables, illustrations, and photographs.

The Model Q-1212C Pulser is a single frequency (60 pps) generator whose output waveform is characterized by a rise-time of less than 4 mµsec and, depending upon the method of termination, an exponential decay having a time constant of 300 or 600µsec. The waveform approximates that produced by a radiation detector. The waveform approximates that produced by a radiation detector. The maximum available output from the instrument is 10 volts, positive or negative polarity, and is continuously adjustable by means of step switches and a 10-turn potentiometer. The potentiometer has a linearity of 0.1%.

A program has been coded for the ORACLE which will solve, using Monte Carlo technique, the transport problem for monodirectional, monoenergetic gamma radiation incident at an angle Θ, upon an infinite laminated slab of finite thickness. Each of the laminations (or regions) is itself an infinite, homogeneous slab of finite thickness. The code is designed to give estimates of energy deposition, energy flux, tissue dose rate, reflected and transmitted energy current, and the angular and energy distribution of the reflected and transmitted energy current. All the answers except for energy deposition and reflected and transmitted energy current are optional.

An approximation is derived for the time constant which characterizes the rate of energy loss of fast ions moving through a plasma. Using particle and energy-balance equations a simple approximate criterion is derived for the estimation of the importance of energy degradation during plasma buildup in a DCX type machine. Next, there is derived the steady-state ion energy distribution for a case in which energy losses are to electrons at a given temperature and particle losses are by charge exchange. The distribution function is used to compute loss rate, upper critical current, ionization rate, and other functions of interest. Quantitative application is made to DCX-2 under various conditions of operation of carbon and deuterium arcs.

Program using the IBM-704 computer to calculate certain kinematic quantities for any non-relativistic reaction of the form.

The effect of adding small quantities of fuel or poison to the HRT has been estimated using perturbation theory. The results have been reduced to a single relation and a set of graphs which make the estimation of added reactivity relatively simple. The perturbation theory results are compared with multigroup results and reasonable agreement is demonstrated; however, there is some question concerning the prompt neutron lifetime.

Report documenting ongoing research and development undertaken by the Metallurgy Division of the Oak Ridge National Laboratory.

Report documenting ongoing experiments, designs, and tests undertaken by the Oak Ridge National Laboratory for the Molten-Salt Reactor Project.

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

A mathematical model is presented for the prediction of heat transfer coefficients for fully developed turbulent flow of fluids in circular pipes by analogy to the transfer of momentum. There is also presented an empirical velocity distribution equation derived from existing experimental data for use in the analogy model. Heat transfer coefficients for fluids with Prandtl numbers ranging from 0.01 to 100 and Reynolds numbers ranging from 5x10^3 to 10^7 are presented in tabular and graphical forms for both the case of constant heat flux at the pipe wall and the case of constant temperature at the pipe wall. The heat transfer coefficients computer in this investigation are compared with existing experimental dat, and a discussion of the parameters affecting the heat transfer characteristics of fluids in turbulent motion in circular pipes is presented.

Report documenting research and developments made by the Health Physics Division of the Oak Ridge National Laboratory.