This effort is being directed toward support studies for the national endeavor on in situ coal gasification. This task involves the investigation of reaction-controlling variables and product distributions for the gasification of both coals and chars utilizing steam and oxygen. Included in this task is the investigation of the effects of using brackish water as the water supply. The high-pressure char gasification system has been received from the manufacturer and is currently undergoing testing. The types of experiments that would be most useful in their studies have been discussed with two of the three laboratories carrying out field tests of in-situ gasification.

This work is being done in support of the development of processes for converting coal to liquid fuel of low sulfur content, suitable for use in power production. Most of the effort is intended to produce information applicable to the SYNTHOIL Process. In the SYNTHOIL Process for converting coal to a low-sulfur fuel oil, coal is liquefied and hydro-desulfurized in a turbulent-flow, catalytic packed-bed reactor. A slurry of coal in recycled oil is reacted with hydrogen at 450 degrees C and 2,000 to 4,000 psi in the presence of Co-MoSiO2-Al2O3 catalyst. The turbulent flow of fluid prevents the coal's mineral matter from settling and plugging the reactor. The gross liquid products are centrifuged to remove the unreacted solids. The centrifuged liquid product is a low-S, low-ash fuel. The following four tasks are included: (1) heat of reaction of hydrogen with coal slurries; (2) heat transfer coefficient; (3) additives to facilitate separation of solids from liquids; and (4) catalyst testing. These are now in the planning stage of development.

Report of equations for calculating decay of U235. The introduction states" Calculations have been made on the U235 fission product decay schemes. The results for a typical example, that of a reactor operating at 1000 kilowatts for 180 days, have been tabulated and graphed. General formulae have been used so that the results can be applied for any power level and any time of irradiation" (p. 2).

Report describing an Isotope Production Reactor, which may be of interest to industrial groups studying the possibilities of entering the radioactive isotopes production field. Detailed drawings and specifications for the construction of the reactor are included.

Report determining the solution of the transport equation to determine accurate conditions for the criticality of a two-region fuel-moderator assembly with reflecting boundaries.

Report describing a pilot plant constructed at Argonne National Laboratory for studying two major process steps for the recovery of uranium and plutonium from spent nuclear fuels of power reactors. A major objective is the demonstration of optimum process conditions for the two steps for synthetic reactor fuel compositions, including those containing mixtures of inactive fission products.

Equations governing the variation of U235. U238, Pu239, Pu240, and Pu241 have been derived and their solutions plotted as a function of irradiation time. The initial U235 content of the uranium was varied from 0.5% to 2.0%. The range of conversion ratios was from 0.5 to 1.2. The irradiation was from 0 to 20,000 mwd/ton of fuel. Since a range of initial conversion ratios is associated with each value of enrichment, a solution results in a family of curves for each isotope, and, since the range of enrichments is large, the number of curves is quite large. Translation of the isotope curves to reactivity variation necessitates a calculation requiring a modest amount of time for a particular case but a prohibitive amount of time to cover the entire range of possible combinations of enrichment and initial conversion ratios. Reactivity variation as a function of irradiation time has been computed for a natural uranium reactor with an initial conversion ratio ranging from 0.7 to 1.2 and for 3 types of reactors in which there is a considerable current interest. Similar calculations for other reactors can be made by making use of the isotope curves and the calculation technique set forth in …

The production of cross-type control rods having a span of 4-7/8 in., and an arm thickness of 1/8 in., was studied. Extrusion techniques were developed for producing cross-type control rods from each of two alloys; one containing 75% silver-25% cadmium, and the other containing 67% silver-30% cadmium-3% copper. Fabrication of the extruded crosses into clad control rods for the Mark I naval reactor was attempted. A set of unclad control rods for the Zero Power Reactor was produced. The effect of copper, nickel, aluminum, palladium, and indium, singly and in various combinations, on the physical and mechanical properties of silver cadmium was studied. Data are given on the work hardening and annealing of binary silver-cadmium alloys, and on the precipitation hardening of certain complex silver cadmium alloys. A materials specification and suggested fabrication procedure were established for nickel-clad extruded silver-cadmium control rods.

Report issued by the Argonne National Laboratory covering the quarterly report from the Reactor Engineering Division. A summary of reactor programs, designs, development, and experiments are presented. This report includes tables, illustrations, and photographs.

Progress is reported on (1) direct cycle boiling reactor studies, (2) solvent extraction, (3) fluoride volatilization separation process, (4) elevated temperature separations, (5) fluidization studies, (6) development of analytical techniques, (7) processing and utilization of radioactive wastes.

Experiments performed by the Laboratory with the Borax Reactor at the National Reactor Testing Station have demonstrated that a boiling reactor possesses inherent safety characteristics which have not previously been included in the estimation of reactor hazards. Other operating characteristics of Borax were also sufficiently attractive to justify the development of boiling reactors for package power and central station power plant applications. Accordingly, a proposal was made to the Atomic Energy Commission that Argonne design, construct and operate a pilot-scale boiling reactor (BER) as part of the Commission's five year program for development of power reactors. Tentative approval for this project has been granted. The primary objective of the BER is to establish the feasibility of operating a boiling reactor in conjunction with a turbine generator on a scale which can be extrapolated to large sizes. A preliminary evaluation of hazards is hereby submitted for the purpose of determining site requirements for a 20 mw reactor of this type. Because the construction of the reactor would be expedited and its usefulness as an operating experiment greatly enhanced, it is suggested that the reactor should be constructed at the DuPage site of the Laboratory. If the inherent features of safety of …

This quarterly report discusses ongoing research and experiments at Oak Ridge National Laboratory in the Metallurgy Division. This report discusses water cooled reactors, liquid metal cooled reactors, reactor development metallurgy, basic metallurgy, applied metallurgy, and aqueous corrosion.

Theoretical and experimental investigations of the use of soluble poisons (neutron absorbers) to supplement mechanical control rods are summarized. Experimental evaluation of poisons of interest includes in-pile and out-of-pile tests simulating anticipated reactor operating conditions. Other phases of the investigation included methods of poison injection, removal, and cleanup of poison-diluted reactor systems, as well as studies to evaluate possible application of soluble poison control in existing and proposed reactors.

Thermal cycling tests on uranium have shown that the dimensional changes that occur on cycling in the alpha range are directly related to both the texture of the material and its grain size: cold rolled rods generally elongate in the direction of rolling, while the same rods, after a beta-treatment, grow at rates several orders of magnitude lower. This considerable improvement by beta-treatment has been attributed to the texture randomization accompanying the heat-treatment. In the course of this heat-treatment, however, considerable grain growth occurs, which ahs the effect of causing surface roughening on cycling (also referred to as "bumping"); fine grained material generally retains a smooth surface. These observations led to the speculation that the most desirable structure in uranium, from standpoint of dimensional stability, is one that combines both a random texture and a fine grain size. Heat treatment of rolled rod offered no easy method to obtain such a product; powder metallurgical techniques, however, appeared ideally suited for the purpose. To this end, early in 1949, the Sylvania Electric Products Company initiated a program to develop suitable techniques for producing uranium powder compacts having the above-mentioned desired characteristics. Because of the availability of thermal cycling equipment at Argonne, …

It was demonstrated by differential thermal analysis (DTA) that: 1. Catastrophic amounts of energy can be stored in diamond. 2. Even at low irradiations, the release takes place over serval hundred degrees, indicating a spectrum of activation energies. 3. At higher irradiations, the stored energy release is considerably less than the increased energy contents and seems not to have been completely released even at the highest temperatures reached. 4. There is some indication of an increased heat capacity below the temperature of stored energy release. It was shown by DTA that large amounts of energy can be stored in silicon carbide on irradiation. The release was found to be spread out over a greater range of temperatures than in diamond and indicated a larger and higher group of activation energies. Catastrophic release was not achieved. The amount of stored energy released over the range of temperatures used was 140 cal/g in a sample irradiated in a water-cooled test hole at HEW for an exposure of 265 Mwd/aT.

Progress is reported on (1) experimental breeder reactor program, (2) solvent extraction, (3) fluoride volatilization separation process, (4) elevated temperature separations, (5) denitration of uranyl nitrate in a fluidized bed, (6) development of analytical techniques, (7) processing and utilization of radioactive wastes.

Deposits of corrosion products form on heat transfer surfaces and in radiation flux zones at temperatures around 500F in stainless steel systems operating with circulating water. The report considers the possible harmful effects of such deposits on heat transfer and fluid flow, as well as factors involved in the origin of these corrosion products and in the mechanisms of deposition. The prevention of deposition by chemical, mechanical, and electrostatic methods is discussed.

Concurrently with the production of canned uranium slugs for pile operation there arises the problem of nondestructive testing so that no slug which may fail structurally during operation be placed in the pile. The ultimate goal of any such testing program is to devise nondestructive testing methods which will eliminate defective slugs. A secondary goal of the testing program is to learn as much as possible about the construction of the canned slug so that the mechanisms of failure can be understood. Radiography, an increasingly useful nondestructive test method, offered one possible way of investigating this area.

This quarterly report discusses ongoing research and experiments at Oak Ridge National Laboratory in the Metallurgy Division. This report discusses water cooled reactors, liquid metal cooled reactors, reactor development metallurgy, basic metallurgy, applied metallurgy, and aqueous corrosion.

This technical report is concerned with a theoretical investigation of the variation of the regeneration factor [gamma] in a Th232 - U235 thermal reactor. The abundances of the significant isotopes in the thorium-uranium cycle have been derived as a function of irradiation time at constant reactor power. The change in [gamma] as a function of irradiation time at constant power was calculated for combinations of enrichment and resonance escape probability considered likely to exist in a thermal reactor. The effect upon [gamma] of the the absorption cross section of 91Pa233 and of the fission products has been shown.

In the a study of plutonium production, the analytical solutions of the various isotope production equations are difficult and time consuming when spatial variation of the flux is considered. In an effort to reduce the time and labor required to determine the integrated production of a given thermal reactor, a graphical method of analysis is presented. The method is based on the assumption that the reactor may be divided into a relatively small number of regions such that the flux in each of these regions may be assumed constant. The production of the given isotope for each region is then read from the appropriate graph. The total production is obtained by adding the productions of all the regions.

In this investigation a method has been developed for production of Zrclad uranium plate based upon previous investigations at Argonne and Oak Ridge National Laboratories, and preliminary in investigations in their present work. The present preliminary investigations included experimental studies of the effect of interface atmosphere upon roll bonding and led to the conclusion that thin layers of air between bonding surfaces could be tolerated.

Uranium alloys containing 3 and 6 w/o niobium have been prepared in which the impurity content is rather low. These alloys show very promising resistance to corrosion by water at 260 and 300 degree C. The corrosion resistance is exhibited in degassed water and in water which is saturated with air or oxygen at room temperature before testing.

An improved continuous ether extractor is described. The modifications include a means of safely disposing of the active raffinate and a means of positivity checking the raffinate for completeness of uranium extraction. The results obtained on synthetic samples and on dissolver solutions are given. This work was undertaken because of a need for the determination of uranium in dissolver solutions with an accuracy of 0.1 per cent. After a review of available methods it was decided that a gravimetric determination would meet the requirements of precision and accuracy.