This technical report constitutes a brief review of the work undertaken, entirely or in part, under the Contract AT(30-1)-1772 with eh United States Atomic Energy Commission. The work published during 1959 is listed at the end of this report and copies of the reprints or reports are attached where available. These investigations have been concerned with the examination of defect production and behavior as it may be observed by high frequency ultrasonic attenuation and velocity (modulus) changes in such materials as quartz, silicon, germanium, alkali halides (especially NaCl and XCl), high purity aluminum, and glass containing boron. The irradiations used are cobalt 60 gamma-ray irradiation and the reactor irradiation obtained in the Brookhaven National Laboratory graphite reactor.

The chief purpose of this contract is the development of ceramic materials with high dielectric constant and high dielectric strength values, and suitable for use of dielectrics in capacitors capable of storing large amounts of energy in small volumes. Work performed during the fifth quarter included (1) Material development and sample preparation; (2) Sample testing; (3) Optical studies of titanates; (4) Fabrication of large ceramics; and (5) Capacitor design.

The chief purpose of this contract is the development of ceramic materials with high dielectric constant and high dielectric strength values, and suitable for use of dielectrics in capacitors capable of storing large amounts of energy in small volumes. Work performed during the fourth quarter included (1) Material development and sample preparation; (2) Sample testing; (3) discussion of ceramic density measurements; (4) Optical studies of titanates; (5) Fabrication of large ceramics; and (6) Capacitor design.

The adsorption of acetic and oleic acids on the surface of thoria was studied by means of infrared spectroscopy. An infrared analysis of the adsorbates before and after adsorption using differential double bean methods indicated that the adsorption of organic acids on thoria surfaces occurs by an esterification reaction between the organic acid and the hydroxylated thoria surface. Ammonia vapor will not adsorb on thoria, indicated the basic character of the hydroxylated surface.

Attempts to secure a high degree of synchronization of cell division of the Krebs mouse ascites carcinoma, by combined use of aminopterin and citrovorum factor, and by use of colchicine and N-acetyl colchinol, were unsuccessful. In the course of the work, a new method for securing smears of ascites tumor cells was out, which we believe will be found useful

This is the thirty-seventh progress report of the Laboratory for Nuclear Science at the Massachusetts Institute of Technology. Progress during the period of June 1, 1955 through August 31, 1955 is reported on: (1) Chemistry of the fission elements group, (2) Nuclear chemistry (inorganic) group, (3) Nuclear chemistry (organic) group, (4) Cosmic ray group, (5) Elementary particle scattering group, (6) neutron physics group, (7) ONR generator group, (8) Radioactivity group, (9) Cyclotron group, (10) Synchrotron group, (11) Theoretical group, (12) Personnel listing.

The compatibility of zirconium diboride, boron carbide, and boron nitride with type 304 stainless steel was evaluated as a function of temperature (1000-1200°C), time (1-3 hr). Appropriate loadings of the boron compounds and stainless steel powder were blended and fashioned into a compact powder metallurgically. Each compact was roll clad into a plate and subsequently heat treated at a temperature equal to the initial sintering temperature. Metallographic examination of the fabricated and heat-treated plates demonstrated that none of the systems were metallurgically stable. The instability was generally manifested by the (1) interaction of the discrete boron compounds with the matrix and (2) precipitation of a hypothetically boron-rich phase throughout the stainless steel matrix material.

Infrared spectra of solid NaXO2(Ac)3, with X=U, Np, Pu, and Am, have been observed. From the symmetric and asymmetric stretching frequencies of the O-X-O groups, approximate X-O force constants have been calculated and were found to decrease in the order kU—O>kNp—O>kPu—O> kAm—O, the respective values being about 0.705, 0.698, 0.675, and 0.612 megadyne/cm. From the cell constants for NaXO2(Ac)3 it is apparent that the X-O bond distance decreases in the same order—RU—O>RNp—O>RPu—O>RAm—O. Thus, a decrease in bond distance appears to be accompanied by a decrease in force constant, probably because the bond, though shortened by contraction of the electron shells of the metal, is weakened by interaction with the extra valence shell electrons.

Calculations indicate that it is impossible, even at high oxygen injection rates, to insure an excess of oxygen in the HRT fuel solution if the bubble letdown rate is more than 1 or 2 liters per minute. If, on the other hand, no bubbles are allowed to form, a reasonable excess oxygen concentration can be maintained with an oxygen injection rate which would not tax the capacity of the off-gas system. The xenon poison will be reduced to less than 2% by liquid letdown alone, and if an iodine absorption bed is installed below the catalytic recombiner, the xenon poison should be less than 1% without any bubble letdown. Therefore, it is recommended that sufficient copper be added to prevent the formation of gas bubbles and that the oxygen injection rate be limited to a value which would permit adequate holdup times in the present charcoal adsorption beds, assuming this quantity is sufficient to meet corrosion requirements.

Factors affecting the results of thoria particle size distribution measurements by sedimentation procedures currently and recently employed are considered. The effects of thoria concentration, solvent, dispersant, thoria properties, and other factors are discussed.

A second test of the mockup of the Zircaloy - stainless steel transition joint as used in the HRT reactor vessel has been completed. The joint and bellows have now received 104 thermal cycles and 148 mechanical deflections. The joint and bellows have functioned properly; corrosion damage has been negligible, except for a small area on the bellows which has undergone pitting attack. Long-term runs with uranyl sulfate solutions of the concentration proposed or use in the HRT have shown the solution to be stable at 300 C. Substituting heavy water for normal water caused no difference in either corrosion or solution stability. Experiments in which chromic acid was used to pretreat stainless steel have shown that, under certain conditions, the pretreated film can exist in uranyl sulfate solutions at flow rates in excess of the critical velocity for relatively long periods of time. The practicability of using titanium inserts in high turbulent areas of stainless steel loops to minimize corrosion has been demonstrated. The corrosiveness of beryllium sulfate solutions containing dissolved uranium trioxide has been determined at 250 and 280 C. Laboratory studies with regard to stress-corrosion cracking have shown that high stressed type 347 stainless steel will crack …

Chloride-induced stress-corrosion cracking has been encountered in the Homogeneous Reactor Test during the preliminary testing. The rector is constructed of austenitic stainless steels. It is unique in that it will operate at 250 to 300 C with an aqueous uranyl sulfate solution fuel containing 200 to 500 ppm of dissolved oxygen. The cracking has occurred in a secondary system used for detecting leaks in the flanged joints of the primary systems and in the grooves of flanges in the primary systems. Tubing used in the leak-detection system was found to be contaminated with chloride introduced during manufacture.

After several failures had occurred in the HRT leak detector system, several lengths of this tubing were removed for metallurgical examination. The tubing was of type 304 stainless steel and was 1/4" in diameter with a 0.065 wall. The tubing had been purchased as three different lots, the first in 45 ft. lengths and the other two as standards lengths. Tubing from the first lot was used primarily for the shield penetration and, therefore, sections of it are present in all lines of the system. It appears that chloride contamination entered the system in a portion of the first lot of tubing used for the shield penetration. The exact source of the chloride cannot be determined, but after considering the results and visiting the manufacturer's plant, it appears most likely the contamination was during the manufacturing process.

Resistance of carbon steel to corrosion in oxygenated high-temperature (250C) water was unexpectedly good at high oxygen concentration. Pertinent literature, critically examined, and toroid experiments indicted that at low oxygen concentration attack did increase with concentration, but as oxygen concentration was sufficiently increased, more protective films were formed on the metal. Some corrosion factors in the application of carbon steel to nuclear reactors systems are discussed.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the second quarter of fiscal year 1962. The data are summarized in Section II. Discussions on fuel performance, fuel environment (water chemistry), problems with loop operations, and the crud deposition program are included.

This report supersedes the similarly titled report dated December 31, 1958. It describes briefly features of the Dresden plant and proposes technical specifications.

The purpose of this report is to present a conceptual design study on a low power electrical and heat generating plant powered by nuclear energy. The nuclear reactor used in this study is the homogeneous circulating fuel type.

A method is described for recovery of uranium from sulfuric acid leach liquors of Congo ores using anion exchange resins. Excellent uranium recoveries were obtained in cyclic tests which indicate uranium capacities of 11 to 26 lbs. of U3O8 per cubic foot of resin per day. Preliminary tests indicate that sulfuric acid may be as effective as hydrochloric acid in providing the required acidity in the eluant.

The investigation of magnesitic-dolomite uranium ores from Union Miniere du Haute Katanga, Belgian Congo, was undertaken in order to examine the possibility of separating the uranium minerals from the carbonate minerals by Heavy Media Separation or froth flotation; thus reducing the sulfuric acid consumption during the uranium leach. In addition, the effect of carbonate leaching the whole ore and carbonate flotation products was examined.

It is proposed to reduce the initial wall hangup and consequent plasma contamination of a stabilized pinched discharge by reducing the Bz pressure difference across the current sheath to nearly zero at early times. Two methods for accomplishing this are proposed, both involving multistage programming of the Bz system.

Chemical and engineering data were obtained for the feed digestion system and the extraction-scrub step of the Monex tributyl phosphate solvent-extraction process for recovering thorium and uranium from nitric acid-digested unclarified monasite sludge. Tests of the recommended conditions in a 2-in.-dia pulsed column demonstrated that thorium losses were approximately 1.2% and uranium losses, 1.5%. The flowsheet is workable but is not necessarily optimum.

A number of questions have arisen in regard to the interpretation of personnel monitoring results reported to users of the Brookhaven neutron monitoring service. The original announcement was rather brief and did not contain most of the technical characteristics upon which an evaluation of results must be based. The following paragraphs have been composed with the hope that they will clarify the meaning of the exposure reports.

The relative neutron capture gamma production in several prospective iron thermal shielding materials for the New Production Reactor was measured to determine the merit of adding boron to the metal. It was found that for the beam geometry the used addition of 1 1/2 percent boron to the iron before casting reduced the soft gamma production by a factor of 6.5 and the hard gamma production by a factor of 10. No attempt was made to measure gamma or neutron transmissions.

Since the calculation of the PRTR energy distribution was first carried out by J. R. Triplett, the design has become sufficiently fixed to allow a refinement of his values. The present analysis, also, includes a calculation of the fraction of energy which is released in the shroud and process tubers that flows to the primary coolant to the top and bottom shield coolant is taken into consideration. Nuclear data used in the original calculations still appears satisfactory and is, therefore, utilized in the present analysis.