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."

From summary: "The influence of the structure of phosphate esters on the distribution coefficient of uranium between solutions of these esters in carbon tetrachloride and aqueous nitrate solutions has been studied. Esters of secondary alcohols showed higher distribution coefficients than primary alcohols. Phenyl esters extracted less uranium than alkyl esters, and the presence of other negative groups also resulted in less extraction. Compounds containing carbon to phosphorus bonds gave higher distribution coefficients than the corresponding esters, and the distribution coefficients increased with the number of such carbon to phosphorus bonds. Tributyl phosphine oxide (with three carbon to phosphorus bonds) showed high distribution coefficients with many salts, including uranyl nitrate, uranyl sulfate, uranyl chloride, thorium nitrate, plutonium (IV) nitrate and some of the fission products in nitrate solution. It is planned to study these and similar compounds further."

From abstract: "On the 86-inch cyclotron a 41-kw beam has been calorimetered at a net ion loading efficiency of 40%; the average proton current was 1.85 ma at 22.5 Mev. Practical specific yields have been determined for (p,2n) reactions on zinc and bismuth. The investigation of products of proton-induced fission of uranium has been continued and new techniques are being used in measuring angular distribution of reaction products. The 63-inch heavy particle cyclotron is now in operation; N+++ particles have been accelerated to ~25 Mev. In preliminary tests, induced activities have been detected in carbon, nitrogen, and oxygen targets. A hot-cathode ion source is ready for test operation. The 22-inch cyclotron is being used in an investigation of the problems associated with the use of rf and dc electrodes for the acceleration of protons from the ion source into the dees. Radiation-induced corrosion in Inconel tubing containing #21 eutectic (ANP) has been produced by proton irradiation in the 86-inch cylclotron; it is shown that the corrosion was not due to thermal effects. Approximately 150 grams of highly purified U 238 (< 5 ppm U 235) have been prepared, and two grams of thorium 230 (ionium) has been enriched to 90.6%."

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.

Progress report of the Oak Ridge National Laboratory Metallurgy Division providing updates on various projects, experiments, and other work. This report includes a summary of nondestructive testing, physical chemistry of corrosion, fundamental physico-metallurgical research, HRP metallurgy, process metallurgy, metallurgical materials and processing, metallography, and ceramics research.

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 …

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.

Quarterly technical report including reports on chemistry of source, fissionable, and structural elements, nuclear chemistry, radio-organic chemistry, chemistry of separations processes, chemical physics, radiation chemistry, and instrumentation of the Chemistry Division of the Oak Ridge National Laboratory (ORNL). [From Abstract]

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.

The system Dovex 21K and aqueous uranyl sulfate solutions were studied in the solution compositions range: 0-0.005 mol/liter uranyl sulfate, 0.15 mil/liter total sulfate, 0.02 mol/liter H2SO4 (pH 2).It is shown that UO3SO4 and/or UO2(SO4)2-/2 are sorbed on the resin to form R2UO2(SO4)2. It is surmized that these same species are also involved at high uranyl concentrations and pH 2 it is suggested that U2O5SO4, U2O5(SO4)2-/2, and R2U2O3SO4 species may be involved.

Some results have been obtained which may be of use in studying the escape of fission fragments from slurry particles which contain fissioning material. Assuming that the fission fragments are of fixed range H, that they are generated uniformly and isotopically in a sphere of diameter D, the escape probability , mean path length, and path length distribution function have been derived.

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 short-term study was made to investigate the removal of radioiodine vapor from air-steam mixtures as could occur immediately following a nuclear incident in a pressurized water reactor. Activated charcoal traps, designed to simulate a small section of a commercial charcoal canister, were tested at gas velocities of 23.9 to 74.9 ft/min over the temperature range of 75 to 118°C. The iodine removal efficiency was found to range from 99.80 to 99.4% with an average of 99.9%. One test performed at a temperature of 105°C and with gas velocity of 290 ft/min indicated that the iodine removal efficiency was reduced to 99.54%. Activated charcoal exhibits a high efficiency for iodine vapor removal under these test conditions and appears suitable for application in the decontamination of air-steam mixtures.

The results of an examination of "O" ring gaskets remove from the HRT following a discovery that the water in the leak detector system contained 1000 ppm chloride are presented.

The essential physical and mechanical properties of materials which enter into the design considerations of high-temperature systems have been outlined, and the relationship of these factors to the detailed design of structural members noted. Particular attention is given to the phenomenon of strain-cycling and relaxation, and several engineering devices wherein these effects can markedly influence design are mentioned. Techniques and apparatus for the measurement of strain-cycling and relaxation properties are described and some recent data for Inconel at 1500 F is reported. A brief discussion is presented on design-criterial for high-temperature applications, and it is pointed out that in some high performance devices the considerations of the strain-cycling phenomenon may impose the design philosophy of "steady-state" operation. Some elementary applications of strain cycling data to design are described, and test experience indicates that analytical techniques can be developed for predicting with reasonable reliability of the fatigue life of structures.

The general solution to the basic differential equation d^4w/dy^4 +4w= -4f(y) is transformed from the primary form treated in most texts to an alternate form in which each integration constant corresponds to one edge condition at y=0. The relationships between the integration constants of the two forms are derived and the values for the transformed functions are tabulated. The particular solution is derived in general and given in unique form for various functional forms of f(y). Matrix notation is used throughout the derivations; however, a knowledge of matrix theory is not need for application of the results.