This report details activities of the Fuels Development Operation for the months of July, August, and September 1958.

The distribution of the heated effluents discharged by Hanford reactors to the Columbia River has been a matter of interest since the early design stage of the first reactors. The pattern of this distribution is a major factor in determining the extent to which a downstream reactor is affected by those upstream, as well as the localized effects on the ecology of the river. Pollutional characteristics of the effluents are three - heat load (or temperature increase), chemical contents and radioactivity. The latter has received the greatest attention in connection with potential personnel exposure and effects on river biota; it has been assumed however, and generally confirmed by sampling that the measure of distribution of any one of these characteristics in the saw an for the others. Observed distributions of radioactivity for various river and reactor flow rates are documented. Unfortunately, any extrapolation of those observed distributions to altered flow conditions of river regimes is of questionable validity. Mathematical models of the problem have been formulated but have been of little value due to the necessity of measuring certain parameters under the conditions for which a solution is desired. Even so, calculated distributions provide only general patterns and would not …

Prior to the installation of improved charging machines and charging techniques serious gouging of the soft aluminum fuel element jackets was not infrequent. A certain amount of mechanical abrasion also occurs as the element is pushed over the tube ribs in the course of the loading operation. To eliminate or at least minimize mechanical damage there has been some experimentation with hard anodic coatings applied to the aluminum fuel element jackets. This report is an evaluation of thinner anodic coatings which should develop lower mechanical stresses within the oxide and should be less subject to cracking.

The development of a matrix method of solving multigroup diffusion equations in nonmultiplying regions is described. The method is applied to a three-region shielding problem, and comparison is made with experimental results. Equations obtained by this technique can be solved with a desk calculator. (auth)

This is the monthly report for the Hanford Laboratories Operation, September, 1958. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology financial activities, biology operation, physics and instrumentation research, employee relations, 4000 program research and development, operations research and synthesis operation, programming, radiation protection, and laboratory auxiliaries operation are discussed.

The assumptions and equations used to conduct reactor plant load trnnsient studies on the analog computer are presented. The study was performed to determine the magnitude of reactor cooling water temperature and volume variations caused by secondary plant load transients, and to establish the size of the pressurizer which would be capable of limiting the cooling water pressure variations caused by the volume surges. (auth)

Four programs written for the IBM 650 to calculate the gamma dose rates in the primary shielding of thermal reactors are described. Their functions are outlined as follows: Program 651-calculates the core attenuation coefficient and equivalent core gamma volumetric source values for a specific core. Program 652- calculates the activation gamma source data in the shield and prepares tabular data in machine storage for Programs 653 and 654. Program 653- calculates the gamma dose rates in the shield due to gammas arising from activation of shield materials. Program 654calculates the gamma dose rates in the shield due to gammas arising in the core. Gamma photo source values are obtained on the basis of two group neutron flux distributions throughout the reactor core and shield. (W.D.M.)

As new designs, geometries, and procedures for fabricating experimental fuel elements are developed, it becomes necessary to determine how well each new product fills the need which prompted its development. While the criteria for satisfactoriness vary with the type of fuel element, there is one criterion that is common to all types designed for use in heterogeneous reactors: the elements must withstand irradiation in the reactor for the required period without jacket rupture or excessive change in dimensions. To learn whether a new type of element meets this and other requirements, a statistically significant number of representative samples must be subjected to, first, comprehensive screening tests designed to insure trouble-free performance in the reactor; next, a long-term irradiation exposure equivalent to or exceeding that which will be normal for that type of element; and finally, a thorough examination to determine the extent of quality loss resulting from the irradiation. This report discusses the application of the foregoing philosophy in the evaluation of two lots of experimental fuel elements fabricated by unconventional procedures.

Safety requirements to be used in establishing standardized safety practices and procedures for engineering design of radiochemical processing facilities are presented. Revision and expansion of future issues are planned. (J.R. D.)