Report issued by the Oak Ridge National Laboratory displaying a single chart showing measurements of the gamma-ray, fast-neutron, and thermal-neutron dose rates.

From abstract: "The extraction of acids with long-chain amines is described. Preliminary results on the extraction of the amine salts of polonium, plutonium and uranium are given."

From abstract: "The variation of the reactivity of a pile as a function of width of a transverse gap is obtained. The method involves first finding the boundary condition satisfied by the flux at the gap face. This, in principle, provides enough information for a complete solution of the pile equations. A method for calculating the reactivity change is presented. The calculated reactivity is compared with experiment and a brief discussion of the validity of the approximations is given."

Technical report describing a rapid and quantitative radiochemical method for the analysis of neptunium (237). This method is based on the extraction of neptunium (IV) with thenoyl-trifluoracetone in xylene. Neptunium (237) is separated free from other alpha emitters and from non-radioactive interferences, and the method may be readily adapted to remote control. [From Abstract]

Technical report outlining the investigation of corrosion on MTR type fuel elements for the Shielding Reactor in filtered water. Report includes two basic types of protection: element pretreatment by either anodizing or alodizing, and solution control using nitric acid to maintain a pH of 5.5 to 6.5 or the addition of 60 ppm sodium chromate as an inhibitor. [From Abstract]

Technical report calculating "effective removal cross sections" for Pb, Fe and O from measurements of fast neutron does in the water surrounding the BSF reactor. The values for Pb and Fe agree quite well with those previously determined from Lid tank data, whereas that for O is somewhat lower. [From Abstract]

Technical report covering classified work of the Critical Experiments Program in the Physics Division for the period December 20, 1950 to March 20, 1951. Report outlines investigations of properties of critical assemblies composed of uranium and various reflectors and moderators, preliminary measurements of the effective energy for fission in the assembly. [From Introduction and Summary]

Report discussing the progress of various research projects by members of the Mathematics Panel at Oak Ridge National Laboratory for the quarter ending April 30, 1952.

This quarterly progress report details the ongoing research happening at the Oak Ridge National Laboratory. In particular, this report discusses the current status of the Homogenous Reactor Experiment, boiling reactor and slurry studies, and general homogenous reactor studies.

Methods for the determination of plutonium and uranium in highly radioactive scrup dissolver solutions have been developed. Plutonium was separated from the dissolver solutions by solvent-extraction and ion-exchange techniques and determined by potentiometric titration. Uranium was separated by ion exchange and determined by potentiometric titration. Solutions that were similar to the actual dissolver solutions and that contained known amounts of plutonium and uranium were analyzed by these methods. Evaluation of the data secured for the determination of plutonium and uranium by the methods given herein indicated that, within the limits of the precision of the methods, there was no bias. The precision of the data obtained for the determination of plutonium, expressed as the relative standard deviation, was better than 2% for plutonium in the concentration range of 0.27 to 0.64 mg/ml. The precision for uranium was estimated to be about 0.2% for uranium concentrations of 425 mg/ml. These methods and the data obtained by then are discussed in this report; the procedures are appended.

A theoretical and experimental investigation of the coefficient for diffusion of ions across a magnetic field Is described. The resultant diffusion coefficient is found to vary inversely as the square of the magnetic field strength, in accord with the usual collison-diffusion theory. The magnitude of the coefficient is much larger (x700) than the coefficient predicted by the usual ambipolar diffusion theory. This discrepancy is resolved by showing that diffusion across a magnetic field is not ambipolar in character in most arc experiments. The final experimental and theoretical values are in good agreement, and it is unecessary to postulate any additional diffusion mechanisms, such as plasma oscillations.

Part I. Experimental Reactors: The effect of prompt-neutron lifetime upon reactor safety was investigated for the HRT. It was found that for a given pressure rise the allowable rate of reactivity addition was relatively insensitive to the average prompt-neutron lifetime, although the rate de creased somewhat with decreasing lifetime for the higher pressure rises. With only source neutrons present and the reactor initially subcritical, the allowable rate was practically independent of the initial value of k£. For a core-pressure rise of 400 psi, the corresponding rate of reactivity addition was about 0.8% per second; for a pressure rise of 4000 psi, the rate was 2.5 to 3.0% per second. Part II. Thorium Breeder Reactor: An economic study of one-region thorium breeder reactors was completed. Where possible, the process characteristics and cost factors were the same as those used previously in studies of two-region-type reactors. The mini mum-cost reactor is about 12 ft in diameter, operating with 260 g of thorium per liter on a chemical processing cycle of about 450 days. The ratio of U232 to U233 produced is approximately 2 x 10~4 VIM in the minimum-cost one-region system, compared with 4 x 10 5 in the two-region system. The …

The development of the reactor layout is continuing. New features that have been incorporated because of stress, fluid flow, or fabricability considerations include an elliptical fuel expansion tank, a rounded dome to enclose the top of the reactor, a newly designed sodium pump impeller, and other related items. Recently completed heat exchanger tests yielded consistent data from which a series of heat exchangers is being designed. The most promising of these will be chosen for the ART.

Two 500-kw fused-fluoride-to-Nak heat exchangers, two 500-kw NaK-to-air radiators, and a 20-tube high-velocity heat exchanger were fabricated for a heat-exchanger development program. A construction procedure, utilizing both inert-arc-welding and high temperature dry-hydrogen brazing, was used successfully on all of the units. The tube-to-header joints were welded and back-brazed; the manifold joints were inert-arc-welded with full penetration; and the tube-to-fin joints were brazed. A detailed description of the fabrication of each type of component is discussed and a cost analysis of the 500-kw units is presented.

Progress report of the Oak Ridge National Laboratory Analytical Chemistry Division providing updates on various projects, experiments, and other work in ionic analyses, analytical instrumentation, radiochemical analyses, activation analyses, spectrochemical analyses, inorganic preparations, optical and electron microscopy.

Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of the rare gases; the halogens, iodine and bromine; and the alkali metals, cesium and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50 percent of the rare gases, 33 percent of the iodine, 9 percent of the cesium and traces of strontium. After 25% burn-up, the cesium value increased to about 60 percent. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2 percent of the iodine, 10 percent of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15 percent burn …

An estimate has been made of the critical fuel concentration in the HRT, taking into account the effect of the core corrosion sample assembly. The estimate is based on a number of previous calculations of critical concentration in an un-poisoned reactor and one calculation of critical concentration as a function of poison level. The makeup of the first core corrosion sample assembly was used in calculating equivalent neutron poisoning effects. Figure 1 shows the estimated critical concentration as a function of temperature with the corrosion sample assembly in place. At 280°C, the assembly raises the critical concentration by 0.6 g U-235/kg D2O. This effect is equivalent to a uniformly distributed poison equal to 4.1% of the fission cross section. The equivalent poison is greater at lower temperatures, where the uranium concentration is lower.

Various nuclear characteristics of two-region spherical homogeneous reactors have been computed in order to provide information for the design of HRE-3. Equilibrium isotope concentrations were established using an ORACLE code, and a two-group model was used to obtain critical concentrations and flux distributions. Breeding ratio is plotted as a function of reactor size, blanket thorium concentration, and other design and operating parameters, and the time required for a demonstration breeding is discussed. Tables of results, including neutron balances, are given for selected reactors. a number or relations are presented for estimating the effects of fission products, copper, corrosion products, H2O, and the core tank on breeding ratio.

The gamma dose rate in the motor region of the HRT fuel circulation pump was measured with the pump scroll full of radioactive solution. Extrapolation of the data to the solution activity expected in the pump under normal operation gives a dose rate well below that which would result in excessive gas production in the stator can within the life of the pump. The above dose rate does not include the effects of fast neutrons from the fuel solution or of the general cell radiation level in the vicinity of the pump. It appears that the possibility of gas production in the stator from the cell background radiation is sufficiently great to warrant the installation of a shield around the outside of the motor end of the fuel circulating pump.

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.

The nuclear safety portion of this report is inclined to ignore the factors by which safety limits can be increased. It makes no mention of the control that can be exercised by limiting the assay of the U-235 being processed in the plant. From some of the previous reports, it is apparent that this plant is not anticipating processing U-235 of assay greater than approximately 20%. At this value, many of the numbers that are presented in the tables could be increased markedly. Rough examination indicates that these values all refer to top product U-235. The general discussion is, however, excellent. The references apparently used are those unclassified references with which we are all familiar and think highly of. We would recommend the inclusion of TID-7016.

The 300A and loop were operated for 2862 hr with thorium oxide slurry at 1500 psi and 280ºC to determine the effects vane inlet and exit geometries on impeller wear, the wear rate of aluminum oxide bearings in this size pump, and the operating characteristics of the loop. The thoris, a 1600*C-fired oxide, had a mean particle size of approximately 2 u. Average circulating slurry concentration was approximately 450 grams of thorium per kilogram of water and average flow rate was approximately 300 gpm.

High heat flux electrical cartridge heaters were tested with direct air cooling under simulated ORR Loop conditions. The cartridges and the heater design were found to be satisfactory. A gas cooled of concentric pipe design utilizing air, water, and air-water mixtures as the coolant was also evaluated and found to be satisfactory.

In connection with the program of the measurement of eta for U233, several critical experiments have been performed by R. Gwin and D. W. Magnuson of ORML with light water solutions of uranyl nitrate (highly enriched in either U233 or U35) in an essentially bare sphere 27 inches in diameter. This report presents the results of two multigroup-diffusion-theory calculations for the above experiments performed by C. B. Mills and associated at Los Alamos. Assumer cross sections, material concentrations detailed neutron balances and a comparison with elementary theory are included. The agreement between the calculated and experimental multiplication constants is excellent for the multigroup calculation but only fair for the elementary calculation. The latter method overestimates the fast leakage so that the computed multiplication constant is less than that found experimentally.