Extraction of HNO3 by triaryl amine was studied by equilibrating equal volumes of aqueous and organic phases at 25C. At HNO3 concentrations of 2 to 10 N the acid in the organic phase in excess of that equivalent in the amine concentration was proportional to the concentration of HNO3 in the equilibrated aqueous phase. other workers report similar results with nitric acid and tri-n-octyl amine in benzene. Zirconium extractions carried out at 10g Zr/1 with 0.35 M TLA nitrate in toluene showed a fourth power dependence of EZR on HNO (aq) over the range 2 to 10N. Maximum distribution ratios calculated from samarium scouting experiments using amines in kerosene were about 5 x 10(-3) for Primene JMT, 10(-4) for TLA, 10(-5) for S-24, and less than zero for DTDA. Distribution rations in the extractions ranged from ERu = 0.12 for 0.35M TLA shaken with an initially new 2N HNO3 solution for 15 minutes. Data on Zr and Ru standardization in TLA solution for spectrophotometric analyses are included.

From the point of view of constituents of a fuel element at temperatures between 2500 and 4500 degree K, the fuel elements can be considered to consist of six types of material: carbon, elements less volatile than carbon, 26 moles of rare gases, 21 moles of alkali metals, 17 moles of alkaline earth metals, and 4 moles of miscellaneous volatile elements. Various processes involving the constituents from 2000 to 45000 degree K are considered. Reactivity gain due to can rupture is discussed.

The maximum temperature in the interior of the fuel element could be greatly reduced by incorporating a liquid between the fuel element and the outer can to increase-heat transfer rates. It is of interest to consider what liquids would be chemically compatible with graphite and the actinide carbides. Elements which melt below 1100 and boil above 1400 deg C that form no stable solid carbides, include Cu, Ga, TI, Ge, Sn, Pb, Sb, Bi, and compounds include GeP, GeS, GaP, Ga/sub 2/S, GaTe, GaAs, SnTe, Sm/sub 3/As/sub 2/, Sb/sub 3/Te/sub 2/, Zn/sub 3/Sb/sub 2/, Zn/sub 3/P/sub 2/, ZnS, ZnTe, and Zn/s ub 3/As/sub 2/. Some of these compounds have equilibrium pressures that might be considered too high, but they may actually vaporize slowly enough because of low vaporization coefficients to make them suitable. There are probably rot enough data nor adequate theories for predicting the rates, and Langmuir type vaporization experiments would be necessary to determine the rates of vaporization of these compounds. The main problem in the use of a heat transfer fluid is that of reaction between the fluid and the actinide carbides. Thermodynamically extensive attack would be expected. However, it may be possible to make the rate …

The quantity and characteristics of fission products in coolant gases due to leaking fuel elements are discussed. It is concluded that the rare gases, the alkali metals, the halides, and Sb may act as permanent gases to a considerable extent. The other fission products are expected to condense out completely on walls or as dust consisting of metals, carbides, and oxides.

This technical report is a summary of unclassified theoretical work on propagation of one-dimensional shock waves and on the propagation of spherical shock waves in gases.

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.

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.

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.

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 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.

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.

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.

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.

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.

A new colorimetric method for the determination of cholesterol has been investigated. It makes use of the relatively stable green color given by a purified digitonin precipitate with the anthrone reagent of Dreywood. Equal precision can be obtained with the new method using 1/10 or less of the quantity of material required for present colorimetric methods.

It is well known that neutron damage generally causes increases in the yield and ultimate strength and a decrease in ductility of a metal. There is a continuing program at HAPO to determine the extent of these changes in Zircaloy-2 as functions of integrated neutron exposure, irradiation temperature, and reactor atmosphere. Three investigations from this program will be described and the results summarized. The first investigation deals with both annealed and cold worked Zircaloy-2 irradiated at approximately 50 C. and the other two investigations deal with annealed Zircaloy-2 irradiated at approximately 100 and 280 C respectively. In each investigation tensile testing was performed at room temperature.

One of the UO2 fuel element designs proposed for use in the PTRT is the nested tubular concepts. This configuration compromises a central fuel todo surrounded by two concentric tubes of fuel (see sketch, appenx I.) . These UO2 shapes are to be jacketed in the Zircaloy and must be separated from each other and the procuresses tyvm vt annular spaces for the passage of coolant. The annuli are established and and maintained by the longitudinal ribs on the outer surface of all three jacketed fuel elements.

At the request of Design Development Operation, various heat transfer and fluid flow problems were examined which are peculiar to the CO2 Gas-Cooled Loop in the PRTR. The results of these calculations are desired primarily to aid in demonstrating the adequacy of the design proposal. In addition, the operational limits of the loop and the consequences of the gas loop installation on the PRT reactor are of interest.

One of the research studies of the Chemical Effluents Technology Operation is the improvement of the method for predicting the capacity of a crib for the retention of wastes. In addition to laboratory work the research was extended a field experiment using a simulated crib fed with a solution containing a radioactive tracer. The purpose is twofold: (1) to check the validity of laboratory findings and (2) to observe several features of the behavior of solutions put to ground in the field.

Uranium dioxide, uranium monocarbide, and uranium mononitride are potentially useful ceramic nuclear fuel materials. This paper reports the results of a determination of the melting points of these materials.

An empirical relation is presented which represents a boundary between no vacuum sparking and possible vacuum sparking. Metal electrodes and r.f. or d.c. voltages are used. The criterion fits several orders of surface gradient, voltage, gap, and frequency. Current due to field emission is considered necessary for sparking but in addition, energetic particles are required to initiate a cascade process which increases the field emission currents to the point of sparking. An elementary cascade process is outlined, but the data upon which it is based is not fully stated.

The Redox solvent extraction process for the separation of uranium and plutonium from their fission products and from each other has operated on an over-all basis very successfully for more than four and one-half years. During this time, there have been occasional periods of high product losses and/or poor separation of fission products from uranium or plutonium as well as periods of poor operability due, among other things, to emulsification and entrainment. The search for the cause of these various process troubles has resulted in almost continuous testing of the process solvent in attempting to determine its contribution, if any, to these difficulties. This report documents the history of the Redox Plant solvent.