Report describing a 6144-channel analyzer designed and built for the purpose of analyzing time-of-flight during studies of slow neutron scattering at Hanford Laboratories. This includes descriptions of the analyzer, its logic and circuits, and its test mode.

Report describing "computer techniques used in the simulation and parameter determination, how the data were prepared for inclusion in the computer simulation, (and) methods used in presenting the results" (p. 3-4).

Report that describes Hanford Laboratories' overall N-Reactor plant simulation model and its use to study the characteristics of the plant, plant operational procedures, and the effects of plant, operator, and control system malfunctions.

Report documenting "the long-term irradiation stability of the graphite used as the moderator in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge" (p. 1) by irradiating capsules at the General Electric Test Reactor. This includes the design and construction of the experiment, experiment procedures, and results of irradiation. Appendices begin on page 83.

Report regarding the calcination of full level Purex and Redox waste carried out "in a remotely operated pilot plant consisting of a radiant-heat spray calciner and a pot calciner and their associated off-gas equipment" (p. 4). This contains details of the calcination, effects on the waste, and difficulties encountered.

Report describing a portable radiation survey instrument, its design, and usage.

Report discussing the thermal conductivity of uranium dioxide. The first part of the report includes the thermal effects on the substance's physical form, while the second part describes the experimental details.

Report describing tests "in which a double chambered irradiation capsule containing NaK (sodium-potassium alloy) in its inner chamber was deliberately perforated to allow NaK (the heat transfer medium) and water to react within the capsule chambers" (p. 2). The report includes descriptions of the materials used in the tests, test procedures, and results.

Report documenting experience gained by operating the spare PRTR Primary Process Pump. This includes descriptions of equipment used, testing procedures, and operating history.

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.

Through the years, considerable effort has been expended in studies of the explosive reactions sometimes observed in the dissolution of uranium-zirconium alloys in nitric acid. It has been shown (1) that such reactions result from the rapid oxidation of finely dived solids released by the preferential dissolution of the metallic matrix. The explosive portion of such solids has been identifies (1) as an intermetallic compound with the approximate composition UZr2. This compound, referred to as the epsilon phase in previous work, has more recently (2) been termed the delta phase. The latter designation will be employed here.

Disposal of liquid radioactive wastes to ground requires knowledge of the area ground-water hydrology. Information is desired on direction and velocity of ground-water flow. Direction of flow can be determined from maps showing contours on the water table surface, such contours being derived from well water elevations. Ground-water velocity is more difficult to ascertain.

Information was desired on applicability of seismic methods of geophysical exploration in helping to better define and refine hydrological and geological factors affecting ground disposal of liquid radioactive wastes. To date most of information on the area geology has been obtained from surface mapping and well drilling, although it has long been recognized that geophysical techniques may greatly assist in defining the geology of the region. Although direct measurements are generally preferred to indirect measurements have proven extremely valuable in many applications. Careful analysis of field data from geophysical measurements, and correlation with known geologic conditions, should provide significant knowledge of earth formations. Geophysical methods do not directly determine depth, type and composition of underground earth materials, but rely rather on detecting differences in physical properties (elastic wave velocity, magnetic susceptibility, density or electrical resistivity) of such earth materials.

Soil chemistry studies have been carried on at Hanford for a number of years in support of the disposal of low and intermediate level liquid wastes to the ground. Equilibrium type experiments were carried out to investigate the mechanisms of the reactions of Sr, Cs, and rare earth ions with soils (1, 3, 4, 5). Experiments were also conducted with simulated and actual wastes in soil columns and an empirical method was developed for prediction of radionuclide breakthrough from ground disposal facilities into ground water on the basis of short, laboratory soil columns and equilibrium experiments (6,7). In this method it is assumed that the shape and position of the breakthrough curve is the same under laboratory conditions as in the larger scale field case. It is also assumed that the "column volume" or amount of soil effectively used for fission product sorption is that directly under a disposal facility without considering any spreading of waste solutions. These assumptions are known to be on the conservative side, but the degree of conservativeness is not known.

Extremely radioactive particles result from the abrasive wheel cutting of irradiated fuel elements in the cut-off cell located in the 327 Building. Due to the use of water during the cutting operation these particles are in suspension and must be removed before the liquid can be discharged to the liquid waste tanks in the 340 building. The filters presently employed in the cell total activity discharged to the 340 Building tanks is attributed to the particles which pass these filters. With the start-up of a new, larger cut-off cell the cutting is expected to increase considerably, thereby increasing the amount of radioactive cuttings requiring disposal. The installation of equipment which would more efficiently remove the suspended particles would (1) reduce the activity level of wastes disposed to ground, (2) reduce the radiation exposure level to drivers who truck the wastes to the 200 Area, and (3) prevent the forming of localized high radiation zones due to the deposition of particles in the waste line in the 327 Building.

At the request of the Hanford Operations Office of the Atomic Energy Commission, a criticality instrumentation meeting was held at Hanford on August 17 and 18, 1959. The purposes of the meeting were: (a) to review types of criticality control and/or alarm instrumentation at each site, their problems, shortcomings and potential obsolescence; (b) to discuss instrumentation in the area if approaching criticality; and (c) to consider for material balance purposes, the feasibility of devising instruments to detect the buildup of plutonium and enriched uranium bearing residues in process systems.

Sinterability studies on mixtures of UO2 and PuO2 powders have been continued. All of the censities from previous sintering experiments have been recalculated utilizing a somewhat different water immersion technique. Since densities of pellete obtained by physical dimensions are inaccurate due to chipping, and water immersion does not account for the open porosity, the pellets were coated with clear krylon before being immersed in water. This effectively sealed the open pores. Plots of density versus composition still indicate a drastic reduction in UO2 sinterability with even the slightest additions of PuO2 and a rapid increase with PuO2 contents greater than 20 weight percent. The following table compares densities of pure UO2 and PuO2 after an eight hour soak time:

The Zircaloy-a clad, swaged UOa, 19-rod cluster fuel element for the PRTR was designed to use Zircaloy-a wire spirally wrapped around the fuel rods as spacing members. Such use of unbonded, Zircaloy-a spacers introduced the possibility of fretting corrosion. This paper reports preliminary irradiation tests conducted to determine whether or not such corrosions occurs in this fuel element design.

The basic process for fabricating a swaged fuel rod is simple, easy to control and inexpensive. A zircaloy tube is filled with uranium dioxide powder, the ends temporarily plugged and the loaded tube is swaged to compact the UO2 powder to the required density. The swaged rod is then cut to length and counterbored and then end cape are welded into each end. After several tests and inspections, nineteen rods which meet the quality standards are assembled into a single fuel element ready for irradiation.

Very little information and data are available on the various physical properties of Zircaloy tubing, especially the high temperature- high pressure failure aspects. As greater use is being made of the low cross-sectional properties of this material in the design and development of reactor components, it was considered desirable to determine the type of failure that would be experienced under conditions of a localized over-heating. This hot-spot could be caused by "cocked" or warped fuel elements coming in direct contact with the process tube, thereby preventing adequate cooling media at the particular location.

One of the UO2 fuel element designs proposed for use in the PTRT is the nested tubular concepts. This configuration compromises a central fuel todo surrounded by two concentric tubes of fuel (see sketch, appenx I.) . These UO2 shapes are to be jacketed in the Zircaloy and must be separated from each other and the procuresses tyvm vt annular spaces for the passage of coolant. The annuli are established and and maintained by the longitudinal ribs on the outer surface of all three jacketed fuel elements.

It is well known that neutron damage generally causes increases in the yield and ultimate strength and a decrease in ductility of a metal. There is a continuing program at HAPO to determine the extent of these changes in Zircaloy-2 as functions of integrated neutron exposure, irradiation temperature, and reactor atmosphere. Three investigations from this program will be described and the results summarized. The first investigation deals with both annealed and cold worked Zircaloy-2 irradiated at approximately 50 C. and the other two investigations deal with annealed Zircaloy-2 irradiated at approximately 100 and 280 C respectively. In each investigation tensile testing was performed at room temperature.

This work was initiated as an evaluation of the application pf rubber finger ring dosimeters to the control of extremity radiation exposure. The use of film dosimeters is a well known general practice throughout the radiation protection profession, however, film radiation exposure measurements associated with plutonium metal handling present a unique problem. The practical measurement of extremity dose depends upon the accuracy and reproducibility of results under the most severe conditions using a dosimeter of minimum size of the radiation exposure to the hands in the rubber finger ring.

Aluminum nitrate solutions are used as salting agents in the redox process and frequently in organic-aqueous extractions in the laboratory. Determination of specific gravity is a convenient means of estimating the concentration of such solutions. However, where an aluminum nitrate solution has been prepared by dissolving aluminum hydrate in nitric acid, the ratio of nitrate to aluminum may intentionally be other than stoichiometric, the solutions being described as acid, neutral, or acid deficient.