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.

This report is a monthly progress report for the 100 Area of the Hanford Reservation, covering work by the plant assistance group and engineering development group. It summarizes activities of these groups in support of the production reactors in this area.

Monthly reports are given for the B, D, DR, F and H piles. And the Engineering Development Group reports on the: Ball 3-X program, thermocouple slug irradiation and related studies, process water study, metallurgical and corrosion studies, analytical research, heat transfer and temperature effect experiment, assistance to pile physics group, miscellaneous development projects, and a job status report from Engineering Development.

This is the monthly 100 areas technical activities report from the engineering division for the month of March 1951. It reports on engineering activities related directly to the different production reactors, and gives progress reports on various engineering projects which are in development by the engineering group.

This is the monthly 100 areas technical activities report for the physics group for the month of July 1951. This group was concerned with pile related studies. Work discussed includes neutron attenuation measurements in pile shielding test facilities, studies of physical properties of shielding materials (concrete), work on a xenon generator and separation facility, further development and shielding work for a neutron spectrometer, continued work on a magnetic spectrometer, and counting equipment. Studies of neutron fluxes from exponential piles, and criticality studies are also discussed.

Activities discussed in this report are as follows: shielding studies; xenon generator; assay of xenon sample counting techniques; crystal diffraction neutron spectrometer; C{sup 12} cross sections measurements; magnetic spectrometer; scintillation spectroscopy; improved pile structure; criticality for the untamped 16 inch spherical reactor; exponential experiments; large scale pile structure; and operational pile physics.

Acenaphtheno is a low-melting organic compound, C{sub 10}H{sub 6}(CH{sub 2}){sub 2}, used as a temperature indicating coating for slugs tested in the Frost Test induction coil. In preparation for this test, slugs are coated by spraying a 12 to 24 per cent by weight solution of acenaphthene in a high vapor pressure solvent on the rotating slug. The solvent evaporate leaving a thin, adherent coating of acenaphthene. This process is completely described in reference 1. To be entirely satisfactory for this use, a solvent must satisfy four requirements. High vapor pressure -- rapid evaporation. Non-flammability. Low toxicity. Carbon tetrachloride, the solvent used at percent, is operationally satisfactory but undersirable because of its toxicity. The purpose of this test is to investigate the suitability of various solvents as substituted for carbon tetrachloride.

A proposal by ``S`` Division that operations be simplified by adding fixed amounts of calcium and iodine in the reduction step as a time saving device has been investigated. Ten test runs in the 410--430 gram range have been compared with twenty normal runs made in this period. The average yields were 98.27% for the normal method and 98.26% for the test runs. No effect on individual yields by the variance in percent calcium and the iodine ratio was observed. Seventy percent of the runs normally fall into the 410--430 gram range. By controlled compositing of the feed batches, it should be possible to increase the percentage of the PuF{sub 4} runs that fall in this range.

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This summary from the analytical research section, May 1, 1951, briefly describes the analytical method to be utilized in the recovery of plutonium from crucibles program.

This report is the descriptive summary Analysis of Heat Transfer, Burnout, Pressure Drop and Density Date for High Pressure Water.

The Pu in one batch (160 gms) is precipitated in a purification hood and waste supernatant is then transferred to the storage tank in one of the recovery hoods. Under normal process conditions a small amount of Pu remains in the supernatant. However, it is possible, though extremely unlikely, that the entire 160 grams would be transferred to the waste storage tank. The supernatants are transferred to an evaporator where they are concentrated. This evaporator has dimensions approximately the same as those of the storage tank, however, as concentration proceeds, the remaining solution can pass to a geometry more favorable for a chain reaction. There should be definite assurance before evaporation is begun that the amount of Pu in the evaporator does not exceed 160 grams. It is concluded that there is no objection, from a nuclear safety standpoint, to the discontinuance of analysis of the supernatants.

Preliminary meetings on coring of the graphite have been held with interested parties of the Technical and Production Divisions at which time the advantages and problems of coring were discussed. The advantages and the problems which must be surmounted in order to core have been presented to the working committee. A study of the means of accomplishing coring has been made considering only the central and maximum fringe coring. It is important that the minimum and maximum coring be established as soon as possible so that the coal coring pattern can be worked out and the design completed. From the studies that have been made thus far the 10 mil center core-60 mil fringe core appears to be the most favorable choice. The advantages and disadvantages of several possible coring combinations (central to fringe) are tabulated. The statements and graphs of temperatures made in this tabulation are based on the results of the analogue tests. The estimates of pile expansion resulting from various coring patterns are based on report HDC-1987 by R.K. Andersen. These results are summarized in Figures 1, 2, and 3. The results of the analysis which Technical plans to make on the DR graphite sample will be …

This report describes results of an investigation to determine whether plutonium nitrates and lanthanum compounds were being entrained on the filter aid and filter block of the Nutsche Filters in Cell 2 of Bldg. 231. Leach samples were analyzed for La, Pu, phosphate, fluoride, Pu{sup 240}, and Am. From the data the authors concluded that the holdup on the Nutsche Filter is due to insoluble plutonium compounds either filtered from solution, or more likely, formed on the filter. These compounds would have a relatively low solubility in 25% nitric acid and would dissolve slowly, thus permitting a gradual buildup of material on the filter which was not detected until leaching with 60% nitric acid was instituted.

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This report is based on studies conducted for the Atomic Energy Commission.

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Annual report of Brookhaven National Laboratory describing the work and findings for the lab's departments during summer 1950-1951.

Quarterly progress report of the unclassified sections at Brookhaven National Laboratory providing updates about research and findings.

This document describes the scope of the C-431-B Reactor Production Facility. In dealing with the broad phases of the project, it includes the Sections ``A`` (Scope Modifications) of the approved Design Criteria, modified to ensure correctness to date. Location of the facility has been set as shown on the site map in HDC-2101, designated site number one. Included in Project C-431-B are the 105-C Building, including within that building facilities previously located in the 1608 Building, a contaminated effluent crib adjacent to 105-C, and gas facilities using the 115-B Building interconnected with 105-C. Also included are an oil shed, a thimble storage cave, a badge house, and an exclusion fence. Building services and process lines will be considered part of the project to a location nominally five feet outside of 105-C.

Recently three H-10 tubes (2,772, 2,176, 3,673) have been removed and the tritium and other gases have been extracted from the Z slugs. It is of interest to compare the calculated volumes of gases with the amounts obtained in these extractions. The comparison is given in this paper. The reasons for the large disagreement are not known at present. As pointed out below, the weakest part of the calculation is the knowledge of the exposure which these Z slugs have received. It is possible that some losses, particularly of tritium, occur during exposure and extraction. In view of these uncertainties, it does not appear likely that more 25 has been burnt up per unit of tritium than has been previously estimated.

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Recent tentative production forecasts indicate that as much as 96.6 Kg plutonium, associated with 190 tons uranium, may be produced each month. The ability of the Separations facilities to process these materials is herein summarized. The Redox Production Plant capacity has been reasonably predicted at about 112 tons uranium per month; this rate assumes only that critical mass control is achieved by the limitation of the possible volume accumulation at any point, and that the product concentration step has been modified to permit the greater capacity. At 600 MWD/T (522 g/t) then, the 112 tons uranium per month capacity limits the plutonium output of the plant to 58.5 Kg per month at 80% operating efficiency. The remaining 38 Kg plutonium per month may be processed in the B and T BiPO{sub 4} plants at an operating efficiency of 89.5%; if this operating efficiency cannot be realized, the postulated production rate may be met by either an increased Redox efficiency or by accepting an increased plutonium waste loss of 1% on some portion of the BiPO{sub 4} production.

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