The increase in process tube outlet water temperature to 90{degree}C resulted in increased emergency water flow requirements as indicated by recently approved curves which establish minimum crossheader pressure requirements for the first three minutes following an electrical power failure. To determine whether 190 Building secondary process water pump capacities are adequate for current minimum reactor water pressure requirements, process water pump trip-cut tests are being conducted at all reactors by the Process Unit in accordance with recommendations of the Reactor Process Committee. The results obtained to data in these tests, as well as current plans for the remaining tests, are presented in this document.

Plant Assistance Group summaries are given for the B, D, F, H, and DR piles. The Engineering Development Group reports on: ball 3-X program, special field instruments, pile flux density measurements, magnesium inlet dummies, assistance to pile physics group, water survey program, boiling studies, slug temperature measurements, and the job status report for engineering development.

Pile Engineering reports on: metal deformation studies, status of group I metal, flow laboratory activities, boroscope inspection details, pile control details, elimination of silica feed, controlled exposure temperature for graphite samples, naval fuel test channel, strength of masonite at elevated temperatures, carbon dioxide pile temperature, H pile graphite temperature, sampling of pile graphite, special pile motion measurements, pile expansion, far side restraining clamps, vortical thimble thermocouples, and boiling studies. The Physical Chemistry Group reports on: pile annealing studies, thermal annealing studies, empty process tube temperature, X- ray line broadening, X-ray diffraction studies, the CSF program, the WSF program, and higher temperature exposures.

A preliminary hybrid 1A flowsheet in which the extraction section operates substantially acid while the scrub section is acid deficient is presented. The effects of added nitric acid on the performance of the acid deficient 1A flowsheet are considered, and the need for control of the entering stream acidities in such flowsheets is evaluated. Data on batch-countercurrent decontamination and plutonium recovery are included. (J.R.D.)

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The carbonate and organic leaching processes for the recovery of U from its ores are outlined. The Excer prccess (ion-exchange conversion and electrolytic reduction) and the Fluorox process (starch-- HF reaction) for the production of UF/sub 4/ from ore concentrate and depleted reactor fuels are described. The fluidized-bed process for UF/sub 4/ production from UO/sub 2/(NO/ sub 3/)/sub 2/ is also described. Methods for improving the reactivity of UO/sub 3/ and mechanical and thermal processes for increasing the density of UF/sub 4/ were investigated. Applications of fluoride volatility prccesses to feed materials are discussed. (C.W.H.)

The liquid high-level radioactive wastes from the separations plant are stored in large underground tanks where radioactive decay of the fission products in storage gives off heat. In the case of the 241-A underground storage tank farm, for Purex wastes, advantage is taken of this heat to self-concentrate the wastes. The present practice is to permit boiling and concentration in the storage tanks. The vapors given off from the boiling wastes are collected in a vapor header and passed through a deentrainment vessel and on to two contact condensers where the vapors are condensed and intermixed with waste cooling water. Samples taken of the waste vapors have shown a considerable amount of cesium{sup 137} present as well as other types of radioactive material carry over from the waste tanks. For this reason the contact condenser effluent is discharged to an underground crib 216-A-8. Underground disposal of the increasing volume of condenser effluent as larger waste volumes are accumulated in the underground tanks presents a critical problem which is further complicated by the desirability to transfer the condensate waste to new disposal facilities near the 200 West area. The intent of this report is to present the scope of the facilities …

>The analytical procedures for wet adsorption methods for surface area measurements are described. (J.S.R.)

This report talks about the Alkaline method for treatment of high radiation level aluminum waste

ABS>The heat capacity of synthetic alpha ferric oxide was determined at 5 to 350 K. The experimental technique is described, and the heat capacity and molal thermodyamic functions are tabulated. The heat capacity vs. temperature is shown graphically. (J.R.D.)

This report discusses procurement, fabrication, and specifications of aluminium components manufactured by Harvey Aluminium Company in Los Angeles, California.

This document contains Restricted Data as defined in the Atomic Energy Act of 1954 and proprietary information of Aerojet-General Corporation.

An analysis of the self-limitation of power excursions by steam production in heterogeneous water-moderated reactors using MTR- type fuel elements seems to give excellent agreement with Borax data on both fuel plate temperature rise and total energy for excursions in which the reactor is initially at saturation temperature. A simple model for steam production is developed from thermodynamic and heat diffusion considerations. In those cases for which the reactor is substantially sub-cooled, it appears that for MTR fuel elements, at least, all the water in the coolant channels at the hottest part of the reactor must be heated to saturation temperature before steam is produced. The theory is used to discuss the safety merits of research reactors which differ in neutron generation time, steam void coefficients and hydrodynamic characteristics. It is shown that D/sub 2/O reactors of the CP-5 type, because of their long neutron lifetime, can safely selflimit power excursions resulting from much larger reactivities than can the H/sub 2/O reactors of the swimming pool type. On the other hand, the response of the two reactor types to equivalent accidents, such as the addition of a given amount of fissionsble material, is practically the same. (auth)

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A calculation, coded for a fast computer, is discussed that can fairly accurately match the experimental resuIts obtained from the LASL Lady Godiva prompt burst program. In particular, the code can predict the reactivity where the burst yield deviates from a linear dependence on reactivity and commences to rise much more rapidly. The burst width and maximum rates can be predictcd with reasonable accuracy. The calculation is applied to other simple systems, with the result that there appear to be general rules that one can use to relate the bursts of thcse systems. The calculation can predict what fraction of the energy of the system is converted into kinetic energy. It appears that step-function increases of reactivity of about 1O cents or less in homogeneous metal systems develop only a trivial explosive energy. The method was extended to hypothetical accidents in very idealized reactors. Two general models of reactor cores were considered. These are (1) a core of low density in which the characteristic o a threshold is important, and (2) a model in which the spherical shells are alternately normal density and void. The numbers quoted should be regarded only as establishing upper and lower limits. (auth)

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The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

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