This report discusses the operation of the Hanford Reservation`s 300 Area for December 11 through December 17, 1945. They discuss: extrusion, canning, slug testing, a magnesium chloride flux, and the composition changes of a aluminium-silicon dip bath.

None

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1951. 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.

None

The corrosion rates of slug jackets exposed in the Hanford piles are monitored to determine optimum operating conditions and to evaluate proposed changes in process specifications. Selected process tubes are charged with weighed slugs, and the coolant flow rate and temperature rise through the tubes are recorded until the slugs are discharged. The corrosion rates of these control slugs are computed from weight loss data after discharge. The data obtained in this manner are used to find the relationship between slug jacket corrosion rates and the various conditions of power and temperature that are encountered in the piles. This report covers all of the corrosion data obtained to date for slugs that have been cooled with the currently specified process water, which is alum-coagulated and contains 2 ppm sodium dichromate as an inhibitor (see Appendix II for the water composition). All of the slugs were normal production pieces canned in 2-S aluminum.

None

The November 1956 monthly report for the Chemical Processing Department of Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed was the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operations. (MB)

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operations research, inventions, visits, and personnel status are discussed. This report is for November, 1956.

It is believed that the overall safety of reactor operation will be improved upon completion of projects CG-558, 600, even at the higher power levels anticipated. Installation of sub-critical monitoring instruments concurrent with these projects is a factor in this conclusion. Higher power levels will not of themselves increase the probability of a disaster initiating event; however, higher power levels will reduce the time available for remedial action and will increase the severity of the consequences of a disaster. Loss of process cooling water will precede or accompany a reactor disaster. A reactor containing a normal inventory of fission products will surely be destroyed, with release of some fission products, if cooling water is lost at operational power levels or within hours after shutting down the reactor from operational levels. A power excursion along with the water loss, unless causing a puff release of fission product, will only hasten the destruction. A power excursion not caused by loss of cooling water is possible, but appears to be of almost negligibly small probability. Such an excursion will not become disastrous unless a significant fraction of the cooling water is boiled out of the reactor. The scope of projects CG-558 and CG-600, …

The radiators designated Cambridge #1 and #2 were visually examined and modified before operation. The initial examination revealed poor quality braze joints at the top and bottom plates of Cambridge #1. However, the same plates of Cambridge #2 were adequately brazed. The copper core of the fins was exposed in many of the tube-to-fin joints on both radiators, indicating that the braze material had not covered the fin collars. Samples were removed from each radiator and examined for evidence of mass transfer. Samples were also removed from the air inlet and outlet banks of both radiators and examined for fin-to-tube integrity.

The November 1959 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

BS>Some of the fundamental relationships in a nuclear rocket engine are discussed. The equations required to calculate the performance of the rocket are presented. The problems associated with these calculations are also pointed out. (auth)

The C Reactor was proposed for producing tritium. To evaluate the performance of enriched U-Al J elements and natural Li-Al alloy target (N) elements, 60 charges containing both J and N pieces were irradiated under a variety of conditions in C Reactor. No ruptures were sustained; however Tube 3276-C was discharged because of a suspect. Corrosion rates of J elements were not worse than for natural U irradiated under same conditions. Differences between corrosion of J elements prepared by three different methods were not significant.

In accordance with the NPR technical criteria which require that provisions be made for useful isotope production in the control rods, both isotope-producing and non-production rods have been designed. Design Analysis has been requested by Reactor Plant Design to specify the number and location of the isotope rods for the initial installation. This choice, however, cannot be made without knowledge of the fuel element characteristics and a prediction of the reactor operating techniques. Some of the factors affecting the use of isotope-producing rods are discussed in this letter, and the following recommendation is made.

This report, from the Chemical Processing Department at HAPO for November 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in October 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth)

Studies were made to determine revised moderator coolant flow requirements in the EGCR core for the latest design. The temperature distribution in the graphite columns was also determined. The total moderator coolant flow was calculated to be 24,024 lb/hr and 31,020 lb/hr for full power and maximum anticipated power, respectively. It was concluded that the maximum moderator temperature (1220 deg F average over the cross section), at full power operation, occurs near the peak heat flux at a location 6.25 to 7.25 ft from the bottom of the active core. The temperature in most of the graphite columns varies from 555 to 1150 deg F over the lower half of the column and 1150 to 990 deg F over the top half of the column. The maximum surface temperature is less than 1110 deg F. Temperature distributions are shown on graphs. (M.C.G.)

This report for November 1961, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

Ring-joint flanges were studied in thermal-cycle tests as part of the development work associated with Homogeneous Reactor Experiment No. 2 (HRE-2). The purpose of this study was to provide criteria for design, installation, and operation of joints that would remain leaktight under reactor operating temperatures and pressures. Joints ranging from 1/2 in., l500 lb to 4 in., 2500 lb and with various initial bolt loadings were cycled between room temperature and 636 deg F. It was demonstrated that when joints were made up to HRE-2 standards and specifications, leak rates of less than 0.25 x 10/sup -3/ g of water per day per inch of gasket pitch diameter could be routinely- attained. Undamaged gaskets could be reinstalled or new gaskets used with equal probability of achieving acceptable leak rates. The system installed in HRE-2 was provided with a high-pressure buffer system to ensure that the small amount of leakage to the cell would be nonradioactive. (auth)

Development activities on the large-scale, hightemperature, thoria slurry facility at ORNL were suspended in July 1960. The latest design philosophy, the status of component development, operational procedures, and limitations of each section of the system are presented. The design of the facility would be adequate for a singleregion slurry-reactor experiment on a slurry blanket system with a properly developed blanket flow pattern. The equipment consists of a 330-liter circulating system, including a 32-in.-ID pressure vessel designed for service at 2000 psi and 300 deg C, a 3000-lb- capacity low-pressure system, and the interconnecting lines. Flow sheets are included and materials of construction and recommended instrumentation are discussed. The first five experimental runs in the system, which were made to determine the feasibility of operating with fertile thoria slurry in the blanket region of Homogeneous Reactor Experiment No. 2, either as-built or after minor modification of the original system are summarized. The first run, a heat exchanger test run made at 1800 psi and 300 deg C, was considered inconclusive. Repeated plugging was experienced, but this occurred in the connecting lines rather than in the heat exchanger. The next four runs were investigations of the high-pressure blanket system, with the pressure …

Historical developments of the ORNL dosimeter and the ORNL badge meter, Model II, are discussed. Beta-gamma and neutron dosimetry techniques are described. ORNL film dosimetry experiments are summarized. (M.C.G.)

Tables of electronic radial functions are presented. The finite nuclear size effects and the finite deBroglie wave length effects were incorporated in this work. A brief summary of the formulation of the problem is given. (M.C.G.)

This report, from the Chemical Processing Department at HAPO for November 1962, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and weapons manufacturing operation.

None