"The following is an alphabetical listing of terms and units in common usage at Hanford Atomic Products Operation. Other terms and definitions may be added as the need becomes apparent or as revisions or additional terms are issued by the NCRP and ICRP."

Introduction: "This eleventh quarterly report of the research and development activities of the Radiological Sciences Department, Hanford Works, as before, includes some items charged to control but included for general interest. Such are identified as "not charged to research". The previous report of this series was numbered: HW-24131 - April 16, 1952."

Summary: "Expressions for the fast and thermal flux in the core of an externally driven, symmetrical infinite slab reactor with a reflector are derived as a function of the reflector thickness. The driver is represented by an inward current of fast neutrons through the outer boundary of the reflector. The flux distribution curve is plotted for several values of reflector thickness. The object is to determine the effect of changing reflector thickness on the flattening of the flux inside of the core of such a reactor."

This report summarizes the research and development work carried out during December, 1959, and January and February, 1960, for Budget Activity 2790 - Separations Development for Non-Production Reactors. The major effort on Activity 2790 has been completed. Current efforts on the remaining problem areas will enable Hanford to begin reprocessing fuel elements from power reactors which employ depleted or slightly enriched uranium fuels in July, 1962.

Critical mass experiments to be conducted at HAPO will require dilute and concentrated plutonium solutions. Since very high plutonium concentrations (~2,000 g/l) cannot be attained in pumpable solutions, solid plutonium compounds will be utilized for this phase of the work. Mixtures of plutonium oxide and polyethylene, methyl methacrylate or paraffin compacted in containers have been considers for use in highly-concentrated plutonium systems of known hydrogen to plutonium atomic ratio (H/Pu).

This document will be concerned with the design, operation and maintenance of an automatic fuel element stamping machine. In order to take advantage of automatic unloading and length checking at the Acme-Gridley lathes in the 313 Building, automatic stamping of fuel elements is desirable. This, along with the need for a permanent canning line identification, resulted in a request for a machine to automatically stamp HAPO fuel elements on a production basis.

During normal operation the NPR will function as a closed system and the coolant will be maintained at a high degree of purity. Periodically, however, the system will have to be opened for fuel element discharge and/or decontamination. During these periods the reactor will be operated on single pass cooling. The use of deionized and deoxygenated water as the single pass coolant is not economically feasible due to the large quantities of coolant that will be required. At present it appears that filtered water will be used for this purpose. Since filtered water contains substantial amounts of dissolved solids (approximately 100 parts per million), and since it is saturated with air, this type of operation will have to be investigated with regard to corrosion problems. A particularly serious problem will be present after the system has been decontaminated, since the high temperature oxide films will have been removed and the bars metal surfaces will be exposed to the coolant.

Considerable work has been done on decontamination procedures for both corrosion products and fission products. Testing of promising procedures under typical rupture conditions is an important phase of this work. The Irradiated Rupture Prototype (IRP) has been used since August in this evaluation work. This document is one of a series reporting these data.

Operating plant experiences in restoring to useful service failed equipment contaminated through exposures to radioactive processes is of international interest to the nuclear industry. In accordance with a request from the Hanford Operations Office of the United States Atomic Energy Commission this report of the various measures taken in rehabilitating for re-use expensive engineered equipment has been compiled by the General Electric Company. This report is a review and summary of the disposition of contaminated major processing equipment at Hanford Atomic Products Operation during 1958-1959 and a listing of 1) equipment buried, 2) equipment stored, to be buried or decontaminated, and 3) equipment decontaminated and returned to service.

A flexible foil data processing program is described. Raw data on foil radioactivity are the basic input information required. Output may consists of relative activities, saturated activities, and/or cadmium ratio and flux spectrum data. A statistical analysis of the data is executed with the direct calculation, and errors estimated for the output data.

One of the problems encountered in the design of the EGCR has been the possibility of bowing in the outer six rods of the seven rod cluster. The power distribution in an outer rod is asymmetrical with respect to the center of the rod. This causes a variation in the cladding temperature, and might result in unven thermal expansion, or bowing.

The Biology Operation, Hanford Laboratories, required plutonium dioxide particles in five different micron and sub-micron size ranges. These were to be supplied as a water suspension in small containers. It was necessary to design equipment, suitable for hood operation, which would give a satisfactory separation of these particles.

Measurements of k∞, f, p, and ∈ have been performed in the PCTR in support of the EGCR Program. The values listed below were obtained for the 21.875-inch cell used in the PCTR measurements. They are for a nonabsorbing (helium or vacuum) atmosphere.

An ex-reactor test stimulating the charging of a tubular fuel element in a Zircaloy-2 process tube showed a problem of severe process tube scratching and galling may occur if metal sliding contact fuel element supports are used. Both the fuel supports and the process tubes were autoclaved. Although the conditions of this test were severe in that the sharp edges rather than a flat surface of the supports were in contact with the process tube, it pointed out the potential problem of galling by point on small area contact supports on heavy fuel elements. This difficulty had not been observed with the lighter tubular or cluster fuel elements used in KER irradiation. In addition to the loss of metal and thinning of the process tube after a number of charge-discharge operations, these scratches may lead to localized corrosion or act as points of stress concentration.

This paper describes the measurement of control rod strength in the PCTR. The first section outlines the theory and method of such a measurement, and the second section presents the results of the measurement of the Experimental Gas Cooled Reactor (EGCR) control rod strength. The reactor parameter actually inferred is the difference in the infinitive multiplication factor for a control rod supercell with and without the control rod in place in the supercell. These measurements are, therefore, very closely related to the measurement of k∞ for an ordinary lattice cell.

At the present time, cerium-144 and premtheium-147 are the two fission product rare earths that appear most promising for use as the heat source in isotopic power units. Under proper conditions, cerium and the trivalent rare earths can be extracted from the Purex fission product waste stream as an insoluble sodium-rare earth double sulfate. A reprecipitation as the double sulfate, dissolution of the hydroxide, serves to give almost complete separation from the corrosion products, inert constituents of the waste, and from most of the fission products. The cerium and the trivalent rare earths must then be separated from each other. In the case of cerium recovery, it is necessary to remove the trivalent rare earths in order to maximize the specific activity of the cerium. If promethium is the desired product, a preliminary cerium separation is desirable to protect the ion-exchange resin (used for separating promethium from its adjacent rare earths) from the intense high-energy radiation from cerium.

One of the problems involved in the design of NPR is the method to be used for decontamination, Considerable effort has been made in both laboratories and pilot plants in developing and evaluating decontamination processes under ideal conditions. Promising processes must be tested under prototypical conditions of velocity, water quality, manner of make-up, type and degree of contamination, addition of decontaminating solutions, and so on.

This document reports the circumstances surrounding the failure of sheathed thermocouples during a test being conducted in the KAPL-120 Loop. The report is prepared to provide a record of these events and to acquaint those who use these thermocouples for testing of other applications with the difficulties that are sometimes encountered.

High speed data handling and computing systems are coming into increasing use in the automatic control of complex industrial processes. One of the next significant developments in reactor design is believed to be the application of high speed data handling and control systems to nuclear reactors. This document is a study of such a system intended for use with the Plutonium Recycle Test Reactor.

Thermal oxidation of graphite in flowing CO2 is being studied at 650 to 850 C, in a single-pass gas system at atmospheric pressure, by observing weight loss rates. The method is used to provide comparative data for candidate reactor graphites. The effects on oxidation rates of graphite purity, structure, coke type, graphitization temperatures and other manufacturing variables are determined. In addition, the effects of gas flow rates and graphite surface to volume ratios are observed.

A coolant line rupture during operation of a high temperature gas cooled graphite moderated reactor would present a serious hazard. The reactor would immediately depressurize and a great deal of air would be introduced into the coolant stream. As the air passed over the graphite moderator a runaway oxidation reaction would probably ensue unless an adequate safety system were available. This investigation was designed to evaluate chlorine as a reactor safeguard to be used to control a runaway reaction. Throughout this study, a small amount of chlorine in an air stream has demonstrated the ability to substantially reduce the oxidation rate of graphite. This has been the case even where the principal oxidizing agent was molecular oxygen or ozone.

Four capsules containing Al-1.65 w/o Pu and Al-12 w/o Si-1.65 w/o Pu were charged into the MTR. These capsules will be irradiated to a burnup of 80-100% of the plutonium atoms to determine the stability of the material at high exposures. An additional sixteen capsules containing 5 to 20 w/o Pu in Al and Al-Si have been prepared. Eight are awaiting reactor space and should be charged soon. The remaining eight are being recanned to increase the end gap between the core and the can. Also four capsules containing high density UO2-PuO2 pellets canned in Zircaloy are awaiting assignment of reactor space.

Although the corrosion of resistance of Zircaloy--2 clad fuel elements in high temperature water is excellent, some concern has been expressed in regard to the resistance against fretting corrosion. Since this form of corrosion can proceed quite rapidly it could induce fuel element failure through penetration of the cladding. Fretting corrosion if it occurred, would take place either where the fuel element contacted the process tube or where portions of the fuel element contacted each other, e. g. wire-wrapping around a cluster of cylindrical fuel rods. This form of corrosion is defined as corrosion occurring at contact areas and as such, fretting of Zircaloy-2 could take place in many reactor concepts. This report details the results of testing performed to induce fretting corrosion of Zircaloy-2 and the direction of the future tests.

This document authorizes the performance in accordance with the specifications noted, the PRTR Critical Tests described herein. The experiments described have the following objectives: