The proposed ORNL Research Reactor is designed to serve as a general purpose research tool delivering a maximum thermal flux of 8x10^13 n/cm2-sec at the initial power level of five megawatts. Operation at power levels up to ten megawatts is proposed for such items as sufficient cooling capacity is available to handle the increased heat load. The reactor will use MTR-type fuel elements and beryllium reflector pieces in a 7 x 9 grid with moderation and cooling provided by forced circulation of demineralized water. The reactor tanks are submerged in a barytes concrete pool, filled with water, which serves as a biological shield. Experimental facilities include two 18" diameter "Engineering Test Facilities" and six 6" diameter beam holes. In addition, access to the core is available through the water of the pool. The result on the surrounding population of release to the atmosphere of a large fraction of the radioactive material in the core has been computed by two methods. It is shown that under certain conditions off-area personnel could be subjected to greater than the maximum permissible exposure. An analysis of the maximum hazard caused by the release of the entire contents of the core to the local watershed …

This memorandum sets forth a recommended uniform basis for designing the ORN shield.This includes design values for power level and emergent radiation, standards values for various material properties, and basic radiation intensities.

An extension of the concepts advanced by Langmuir regarding the nature of the platinum catalyzed oxidation of hydrogen and the application of the resulting theory to the experimental data observed by Ranschoff and Spiewak for an HRE type recombiner indicates that their data are corrected by the dimensionless equation (see report) equally well, with a mean deviation of 3.8 percent. This expression is recommended as a basis for the design of catalytic recombiners. The catalytic combinations is pictured as consisting of two surface chemical mechanisms, one of which is oxygen diffusion controlled, the other hydrogen diffusion regulated, the mechanism "change-over" occurring at that point in the recombiner where the components are arriving at the catalyst surface by diffusion in stoichiometric proportions. The catalyst volume requirements for three two portions of the bed are shown to be (see report). The hydrogen mole fraction at the mechanism "change-over" point is (see report). And the relationship between the two mass transfer coefficients is (see report). Methods for evaluating the necessary transport properties of the ternary system steam-hydrogen-oxygen for carrying out design calculations are summarized, and the new significant parameters are tabulated and plotted to facilitate these calculations. The question of non-uniform velocity profiles …

In scouting tests, a number of metals used in nuclear reactor fuel elements were dissolved by 44.5-48.5-7.0 mole % ZrF4-KF-NaF fused salt at 675°C through which HF was being passed. These included type 304 stainless steel at 4 mils/hr; type 347Nb stainless steel at 7 mils/hr; thorium at 14 mils/hr; nonirradiated uranium at 17 mils/hr; zirconium at 22-35 mils/hr; titanium at 31 mils/hr; and Zircaloy-2 at 22-46 mils/hr. Only small amounts of volatile fission products formed when irradiated uranium was dissolved. Variables that appear to affect the dissolution rate are the composition of the fused fluoride, the fused fluoride temperature, the HF flow rate, the metallurgical characteristics of the material being dissolved, and the presence of other metals. The low dissolution rate of 0.001 mil/hr observed for nickel suggests that it may be suitable as a material of construction for reaction vessels.

A preliminary chemical flowsheet is presented of a fluoride volatility process for recovering and decontaminating uranium from heterogeneous reactor fuels after dissolution in a fused salt. In laboratory work, a gross β decontamination factor of > 10 4 was obtained in the fluorination of a UF4-NaF-ZrF4 melt by passing the product UF6 through NaF at 650°C. The solubility of UF6 in molten NaF-ZrF4 was shown in kinetic studies to cause a lag in the evolution of UF6 from the fluorinator. Corrosion of nickel in the fluorination step appeared to be 2-4 mils/hr during the time that uranium was present. The average corrosion rate over the process as a whole was less than O.4 mil/hr. Earlier studies were reported in ORNL-1709 and 1877.

A summary of the laboratory development program on aqueous uranium slurry fuels for the Homogenous Reactor Project during the period April 1951 through March 1953 is presented. These investigations were devoted primarily to a study of the uranium oxides in aqueous suspensions. It was concluded that U(VI) was most likely to be the stable valence state in such slurry fuels and it was shown that β-UO3·H2O platelet crystals were the stable modification at 250°C. Very pure slurries of β-UO3·H2O platelets, uranium concentration of 250g/liter and average particle size of about 10 μ, had favorable settling rates and could be easily redispersed. Their viscosity and corrosion rate in stainless steel were comparable with those in water. Exposure of these slurries to pile radiation disclosed that radiolytic hydrogen and oxygen gas pressure comparable in magnitude to those of uncatalyzed uranyl sulfate solutions could be expected. Fission products in the irradiated slurries were predominantly associated with the solids. Radiation also tended to promote caking of these solids on the walls of the radiation bombs. Uranyl phosphate and the magnesium uranates were briefly investigated as alternate system but were not found satisfactory. The program was discontinued before the feasibility of uranium slurries for reactor …

Construction of the HRT reactor shield tank was completed, and the inside surfaces were painted. The roof structure for the tank is being assembled in preparation for an acceptance pressure test. Service piping and instrument lines are being installed in the central room area by ORNL craft forces. This work is approximately 50% complete. Fabrication of all temperature system components, except the blanket outer storage tanks, has been completed.

An ion-exchange -- polarographic method was developed for the determination of iron(III) in Homogeneous Reactor Fuels. Copper, which interferes, is removed from the fuel by plating it onto a cadmium coil. Iron is oxidized to iron(III) by potassium permanganate, and the iron(III) is separated from interfering metal ions by ion exchange on a Dowex 1 resin column that is in the sulfate form. The iron(III) in the effluent is determined polarographically in 0.5 M sodium citrate solution as supporting electrolyte. A fairly well defined polarographic wave is obtained for the iron(III) → iron(II) reduction at a half-wave potential of approximately -0.15 v. vs. the S.C.E. The relative standard deviation of the data for 2 µg of iron(III) per ml of solution in the polarographic cell was 6.5%; for 10 µg of iron(III) per ml it was 0.6%.

In order that fairly accurate thermal stress calculations can be made on the ART, it is necessary to have a reasonable picture of the temperature distribution in the reactor. To get the temperature distributions, and to determine cooling requirements in various parts of the reactor, one must know the heat deposition rates due to alpha particles, beta rays, gamma rays, and neutrons in all parts of the reactor. The present report contains only the basic physical data necessary to determine the heat deposition rates due to gamma rays. Neutron fluxes in the core and reflector regions of the ART are to be obtained from two-dimensional multigroup calculations (performed by the Curtiss-Wright Corporation). These fluxes, in conjunction with the neutron absorption cross sections, determine the neutron capture and inelastic scattering rates in the core and in the reflector. The data in this report permit the calculation of the number of gamma rays originating at various energies at every point in the core and reflector.

A general program has been written to plot curves on the Oracle curve plotter. The supplement includes changes to slow down some of the loops and minimize the possibility of read-around errors and changes to handle special cases.

A Lucite model of a multi-stage countercurrent hydroclone solvent extraction apparatus has been constructed and tested with Amsco-water. The diffusivity of Cs 134 tracer in aqueous chloride solution was measured to check the performance of the capillary diffusivity measuring system. The experimental data from four Druhm runs showed that 1/8in. thick graphite liners are usable for reactor temperatures above the boiling point of sodium.

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 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.

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.

The preparation of essentially pure dimethyl selenide for the Isotype Separations Group, Chemistry Division, is described. The compound was prepared by warming a mixture of selenium, sodium hydroxide, sodium formaldehyde sulfoxylate in aqueous solution for several hours at 50°C and then adding methyl iodide. Dimethyl selenide was removed by distillation.

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.

A development program is being carried out in the pilot plant to improve the equipment involved in monitoring, metering and pumping process streams. The equipment under development includes an alpha monitor, a gamma monitor, a waste water monitor, a liquid flow rate meter, a canned rotor pump, a diaphragm pump, air lifts and a pressure tank pump. This report presents the present status of the testing program.

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.

A conductometric titration method described by Goldstein was adapted for use in a remote analytical facility. The results obtained by mean of experiments made prior to this adaptation indicated that methanol is the most satisfactory medium in which to determine excess sulfuric acid in uranyl sulfate solutions that stimulate Homogeneous Reactor type fuel. When methanol is used, the complexation of hydrolyzable ions with sodium fluoride, as described by Pepkowitz, Sabol, and Dustin, is not required.

The theory of mass transfer into a flowing fluid is utilizes to estimate the effect of gas film resistance on overall rates of transfer through a porous septum. The expressions developed for the mass transfer ratio...

Methods for the noncorrosive removal of volatile fission products and UO2 dust from carbon steel and stainless steel have been developed. Procedures for applying these methods to the decontamination of the EGCR charge and service machines are described.

This manual represents the recommendations of the Instrumentation and Controls Division committee on Electrical and Electronic Symbols and Drawing that have been issued to date, and supersedes the previously issued ORNL Electrical Symbols List and CF-58-12-141, Electrical and Electronic Drawing Standards for Wiring and Device Coding and Applications.

A portable shield has been designed, developed, fabricated and shop tested to provide the HRT with a facility for direct dry maintenance operations. It provides temporary replacement for any one of the lower roof plugs and should permit many operations to be performed without flooding the reactor cell with water.

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.