The first step in a proposed processing method for recovery of uranium from graphite-uranium fuels consists of oxidation of the fuel by oxygen to volatilize the carbon. Residue ash from the combustion step can be treated in a variety of ways to recover and purify the uranium. The combustion step may be caried out by contacting the solid fuel in a fixed or moving bed with a stream of oxygen-bearing gas in a tubular or annular reactor. Oxidizing gas may be introduced to the reactor at several points up the reactor and there may be continuous or intermittent addition of fresh fuel and removal of residue ash.

The first man-made nuclear reactor -- or "pile" as it was then called -- was rather hurriedly improvised and operated in a crowded space under the athletic bleachers of Stagg Field at the University of Chicago on December 2, 1942. Just prior to this time, there began the assembly of a group of physicists with an unusual assignment. They were determined that radiation hazards of unprecedented proportions must be coped with successfully in the conduct of reactor programs as planned. Since these physicists were to be concerned with the health of radiation workers, they were called health physicists. There was no formal instruction available to this first group of health physicists and they perforce received training as they felt their way by firsthand experience and by trial and error. Health physics at Oak Ridge National Laboratory from the very beginning has been organized into three principal areas: applied activities, education and training and research.

Grade CGB graphite is a nuclear graphite which is basically an extruded petroleum coke bonded with coal tar pitch. No carbon blacks are used and the low-permeation graphite is finished through a series of impregnations and heat treatments with a final heat treatment of all components to 2800 degrees C. A listing of the results obtained is given in Table 1. The results at 51 degrees C are considered questionable. There was a slight contamination of the 90% Pt 10% Rh-Pt thermocouples at 910 degrees C but it was not sufficient to doubt the validity of the 910 degrees C results. However, the results obtained at 1015 degrees C should be disregarded because of severe thermocouple instabilities. In addition, the electrical resistance of the core heater at 603 degrees C indicated the thermocouples had a -10 to -15 degree error which is sufficient justification to disregard the 605 degrees C data.

The thermophysical properties of Armco iron such as thermal conductivity, electrical resistivity, and Seebeck coefficient have been extensively investigated and reviewed up to 1000 degrees C. Few investigations of such properties have been made on high purity iron. If such a study is made using the same apparatus to determine the properties of two purity levels of iron, then several significant intercomparisons can be made which add meaning to data on a single material. The systemic errors for a single apparatus are the same, therefore comparison of a property of two similar materials is more significant. A comparison of the property changes with temperature and purity can show the effects of impurities on the mechanisms contributing to a property and allows prediction of the properties of iron as a function of purity. For these reasons a study was initiated on the high-purity iron for comparison to Armco iron.

The experimental details, mathematical models, and typical data for a rapid comparative method for thermal conductivity measurements are presented. The method consists of measuring the temperature change of a small silver sphere after it is brought in contact with a small disk-shaped specimen which was initially at ta higher temperature. This temperature change was calibrated in the range of 50 to 400 degrees C by making measurements on samples of know thermal conductivity. The accuracy of this technique was shown to be between than +-10% with a reproducibility of at least +-2.5%. Using known transport mechanisms for heat conduction in solids and the temperature dependency of the electrical conductivity, a means to judiciously extrapolate thermal conductivity data obtained between 50 and 400 degree C to high temperature is presented.

The purpose of this study was to investigate mathematically the concentration decrease due to particle deposition phenomena in highly concentrated monodispersed aerosols (mean particle size less than 1.0 mu) flowing through ventilation ducts. It was found that, from the standpoint of removal, the decrease in concentration due to deposition on duct walls was insignificant; but, when considering contamination on duct walls, the amount deposited, even though small when compared with the amount in the bulk stream, should not be overlooked.

This report is an inventory record of materials which were received and shipped from the Purex Process during this time period. Material accounting includes uranium, uranium-235, plutonium, and other transuranics.

To satisfy reactor coolant supply criteria requires a knowledge of available emergency flow rates from secondary and last-ditch systems. This document presents a method for calculating high-tank flow contribution to the last-ditch coolant system from routine high tank-high velocity flush data. The measured flush flow is converted to equivalent through-reactor flow at 20 C and at the critical time of high-tank takeover from flywheel decay.

Large scale self-support fuel usage at the K Reactors began in mid-1963. With the advent of the new fuel element concept, self-support development work in different areas was required. One such area was that of the supports, or rails, used on the elements. The initial self-support rail design on the elements, in order to meet production commitments, was only slightly modified from previous test models. These rails were identified as ``bridge rails.`` Development of the self-supports has continued and the first major modification, the ``arch-rails,`` of the original design is now ready for reactor testing.

This document reports the IPD production projections and reactor inventories for Hanford C and K reactors. Tables included contains the following information: Operating efficiency, exposure, conversion ratio, tonnage, production, and power level.

This report presents the dose rates which were observed during the removal of a Zircaloy-2 process tube from C Reactor. The corresponding calculated activity is also presented for the purpose of correlation with the observed dose rate. The tube was removed on July 22, 1964, after having been irradiated by a neutron flux for 560 days (based on time operated efficiency). The tube activity was monitored with a super high range TP (0-5000 R/hr) by suspending a probe through an access port in the floor of the ``D`` machinery room. As tube 3353 was pushed and cut into sections, the highest activity observed was 675 R/hr, which was seen in the center three-foot segment. The corresponding calculated activity was 650 R/hr. The neutron flux profile, as shown by tube activities along its length, was extremely peaked in the center; for example, the area under the activity curve was less than one-third of the area under a cosine curve of the same amplitude and cycle length.

This is the monthly report for the Hanford Laboratories Operation, July 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, applied mathematics, programming operation, and radiation protection are discussed.

A summary of HAPO activities is given to include separate sections on the N-Reactor and other current production reactors. Specific programs and fuel performance for current production reactor fuels is discussed. Also, the production status, fuel performance, development program and process technology for N-Reactor fuels is presented.

Reactor support charge designs and patterns have been numerous and varied over the past few years, the basis for acceptance of a particular support loading being determined from a consideration of fabrication costs and the hydraulic characteristics of the proposed dummy load as it applies to a particular reactor. Since the first three or four dummy pieces downstream of the fuel column (and upstream of the column in a spline tube) are discarded after the irradiation process, it is desirable to use tubular dummies for this portion of the support charge which are less expensive to fabricate and which are easily identified during the charging-discharging process. However, there exists a remote possibility that one or more tubular pieces could be inadvertently placed in a rear nozzle resulting in a flow restriction in the process tube. These arrangements of tubular support pieces in the dummy pattern, i.e., tubulars positioned against the fuel column and/or in the process tube rear nozzles, have been hydraulically tested to determine if there is adequate and unrestricted flow at normal operating header pressures.

The release of uncontrollable quantities of harmful fission products such as I{sub 131} upon processing areas. At present this is done by procedural techniques using cards, filing techniques, and cross checks. These methods are entirely independent of the characteristics of the fuel itself and are subject to failure from human error. It is therefore advantageous to find a method of determining the age of irradiated fuel after discharge from the properties of the fuel. Such a method, with its relative immunity to human error, would be very useful in checking shipments as well as for such jobs as determining the correct age of fuel being stored for accumulation of desirable fission product daughters. It was the purpose of this effort to make a study of possible methods of direct fuel age determination, formulate and perform a test of the most promising method, and analyze the results in terms of practicability here at Hanford.

This report, for July 1964 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; weapons manufacturing operation; and safety and security.

Reference 1 gave mixer recommendations for aluminum spline tubes at the K reactors. On completion of the study, R. F. Corlett requested a more detailed economic study of mixer useage at KW reactor. R. G. Clough also requested additional information on use of mixer elements in fringe tubes. In order to answer these requests, additional information was needed and requested. A. W. Medcalf replied giving enriched metal requirements in support of mixers. No other information vas received so it was not possible to make the economic study requested by R. F. Corlett. This report presents more detail on tube-life estimates associated with use of the eccentric and concentric-wall tube designs in central- and fringe-zones and the results of five downstream thermcouple probes that operated in concentric-wall tubes at KW reactor. Also, modification of the concentric-wall tube (milling down of the ribs) is discussed.

This production test authorizes the irradiation of full length charges of solid aluminum in the KER loops while on single-pass operation with process water cooling. Calculations show that the irradiations will not result in any foreseen reactor or personnel hazards and can be safely performed in the KER facilities.

None

Figures are given.

The purpose of this test was to evaluate the behavior of natural lithium-bearing materials and to provide samples for production and extraction information. Testing of the target elements was done in conjunction with a test evaluating the behavior of NIEl fuel elements. Conditions for operation during the test were similar to, or more severe, than conditions expected in N Reactor.

The AEC has indicated a need for plutonium with a high concentration of the Pu-240 isotope. Concentrations up to 25 W/o Pu-240 have been requested with a firm commitment given for 25 kilograms of plutonium with at least 18 W/o Pu-240. To meet this production demand, irradiation of depleted uranium, 0.22 W/o U-235, will be undertaken. This test authorizes half of the irradiation in a Hanford K reactor, outlines the general plans, and presents the details of the irradiation. The objective of this test is to authorize irradiation of approximately 12 tons of depleted uranium, 0.22 W/o U-235, in a Hanford K reactor. The depleted uranium will be irradiated to exposures which will yield plutonium with at least 18 W/o Pu-240, an average exposure of approximately 1600 MWD/T.

The attached tables are an extension of the data that will be used in the forthcoming Production Studies. It considers thorium load substitution for the natural uranium cases previously reported. The power level, operating efficiency, exposure, conversion ratios, tonnage, and production are given for the natural U loading, blanket loading, and full conversion. The small reactors, the C reactor, and the K reactor are considered.

Investigation of several alternative methods of modifying the C- Reactor overbore test facility in order to terminate the production tests has been completed. The results are presented in this report.