This monthly progress report covers activities in the following areas: pile control; shielding studies; process tube assembly; pile operating equipment; power recovery; pile piping; and test laboratory. A detailed discussion is given for each topic.

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This document describes aspects of heat transfer for the Hanford production reactors. Boiling, flow rate, cooling water failures, thermocouples, and shielding are considered.

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During the preliminary operaton of a volatilization apparatus it became apparent that a large stopcock in the high vacuum side of the oil diffusion pump was quite unsatisfactory. Since it was necessary to mount the stopcock in a relatively inaccessible location, the difficulty of turning the plug and the consequent necessity of regreasing the valve approximately every week were real problems. Also, it became apparent that an increased pumping speed through the valve would be required. Since a larger stopcock was out of the question, a type of valve reported by Stern was modified in design and was bulit into the apparatus.

The question of maximum permissible levels of fission products or corrosion products in the pile effluent water which you raised is not a simple problem that can be completely answered by a list of maximum permissible levels. This water is a complex mixture of various isotopes each of which contributes to the total hazard according to the amount present, its particular metabolic features, and its mode of radioactive decay. Unfortunately, also, application of maximum permissible limits determined for consumption by humans may give only a part of the picture since it is necessary to consider also the effect on the ecology of the Columbia River and the ultimate effect on the aquatic life. With these considerations, we can make estimates of some limits which may be applied in design of new systems with, however, the reservation that practical application may depend on the mixture of isotopes actually encountered in operation. This document provides a discussion of maximum permissible levels of radioactive effluents and corrosion products in the reactor cooling water. 5 tabs.

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The chain reacting conditions for plutonium nitrate in water solution have been examined experimentally for a variety of sizes of spheres and cylinders. The effects on the critical mass of the displacement of hydrogen and the addition of poisons to the fuel were measured in water tamped and bare reactors. In this report the data obtained in the investigation is presented graphically and in tables. Some preliminary analysis has been made yielding the results: (i) the absorption cross-section of Pu{sup 240} is 925 {plus_minus} 200 barns and (ii) the minimum critical mass of Pu{sup 239} in water is 510 grams at concentration of about 33 grams per liter.

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This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

The purpose of this report is to present a basis for the application of slug corrosion rate data to corrosion limitations on pile power levels.

It was the purpose of this study to determine whether it will be more desirable when activatiing Task I of the RMA. Line to revise the existing remotely operated equipment or to remove existing equipment and install a manually operated facility. It was also the purpose of this study to provide sketches and cost estimates for each alternative which could be used as a guide for preparation of a Project Proposal. At present, the process to be used in Task I has not been established. This study, therefore, necessarily covers only processes similar to those for which Task I was originally designed. Information received from the Process Unit indicates adoption of such a process is probable.

The Process Unit has requested the Present Canning group to evaluate two alkaline cleaners, Pennsalt A-10 and Turco ``Aviation`` Cleaner, as possible substitutes for the present trichloroethylene vapor degreaser and Duponol wash. Disadvantages of the present cleaning procedure are: ineffectiveness; expense; toxicity of trichloroethylene. The alkaline cleaners Pennsalt A-10 and Turco ``Activation`` Cleaner were tested and compared with an alkaline cleaner made from Na{sub 2}CO{sub 3} and Na{sub 3}PO{sub 4} and with the present cleaning method. With available equipment Pennsalt A-10 and Turco ``Aviation`` Cleaner did not remove grease and dirt as well as the present cleaning method. A solution of Na{sub 2}PO{sub 4} and Na{sub 2}CO{sub 3} was the best degreaser tested although it did etch the surface of the cap. This document provides a discussion of the procedure utilized and the results.

In 1950, the U.S. Geological Survey studied a uranium deposit, previously exploited, at the Silver Cliff mine near Lusk, Wyoming. The deposit consists primarily of uranophane that occurs as fracture fillings and small replacement pockets in faulted and fractured quartzite of Cambrian (?) age. The country rock in the vicinity of the mine is schist of pre-Cambrian age intruded by pegmatite dikes and unconformably overlain by flat-lying Cambrian (?) quartzite. The mine is at the southern end of the Lusk Dome, a local structure probably related to the Hartville uplift. In the immediate vicinity of the mine, this dome is cut by the Silver Cliff fault, which is a north-trneding high-angle reverse fault approximately 1,200 feet in length with a stratigraphic throw of 70 feet. The fractured quartzite has been the locus for the deposition of uranophane, metatorbernite, pitchblende, calcite, native silver, native copper, chalcocite, azurite, malachite, chrysocolla, and cuprite. The fault was probably mineralized throughout its length, but owing to erosion the mineralized zone is now discontinuous. The principal ore body is about 800 feet long. The width and depth of the mineralized zone are not accurately known but are at least 20 feet and 60 feet, respectively. The …

At the time the first UO{sub 3} from so-called ``full-level`` Redox runs was shipped from 224U, the % U{sub 235} as determined in 222-S appeared high when compared with theoretical values from the burn-out curve. Close agreement between the 222-S results and % U{sub 235} determined by K-25 on lots 007, 008, and 009, however, indicated that a considerable heel of cold uranium in Redox had been blended with the first ``full level`` material. This explanation has been verified by data collected over the past weeks and provided this report. It has also been verified by the % U{sub 235} found in lots later than 010* (after most of cold uranium flushed out of Redox) which ranged from 0.66% to 0.64%.

A preliminary feasibility study is made of seals for a machine running at 8000 rpm, having a 2-in. shaft, and operating with a pressure differential of 5 lb/in{sup2}. The following conditions must be met: during operation the air inleak shall not exceed 50 lb/2000 hr and the loss of Freon 11 shall not exceed 0.02 lb/hr; the leak-up rate during shutdown shall not exceed 1/10 in. of Hg per 24 hr (the contained volume is 100 ft{sup3}). Dry-contact and Freon 11-lubricated contact seals are considered briefly, but neither appears to be practical. The following positive-clearance seals are more feasible and are discussed in greater detail: concentric-cyinder seal, utilization of a journal bearing of the machine as a seal, and a radial-controlled-gap type of seal.

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