A report which covers progress on the testing of the General Engineering and Consulting Laboratory Turbine Pump No. 1 from June 16, 1948, to July 31, 1948.

This report analyzes Parian and Teflon that were exposed to solutions of 1AX, 1AS, and 1AF, ANL June 1, 1948 Flowsheet, under conditions described in the following report.

This report discusses the development of a method for creating leak-proof closures in aluminum sample casings in order to pass a standard bubble-test inspection.

Introduction: A limited test progress on Alloghany-Ludlum type 347 stainless steel plate which failed the Huey test (ASTN A-262-44T) has been completed.

Report detailing the refractive index of solutions of uranyl nitrate, nitric acid and aluminum nitrate. This report builds upon preliminary data and a brief discussion that from a previous report. This report covers more accurate and detailed data that had be accumulated following the initial report.

A preliminary report has shown the potential importance of organic reactor coolants to Hanford technology. The salient points are complete avoidance of corrosion problems and accomplishment of DPR performance with low pressure technique. The recommendation of the report that loop tests be made at Hanford is to be acted upon by Recirculation Technology Unit as soon as organic coolant material is received. To insure starting this experimental work fully abreast of the information available, a trip to sites working with organic coolants was made. Specifically, data were sought on practical details of design for handling organic coolants, heat transfer and fouling characteristics, corrosion of materials, and hazards of handling organics.

The current interest in a tube-in-tube type Zircaloy-2 clad UO2 fuel element for use in the power reactors provided the incentive for a study of the dissolution of cladding from an annular space. The objectives of the study were to compare the estimated rate of dissolution in the annuli with rates on fully exposed metal. Because of an interest from the design standpoint, heat balance data from the study are reported.

Among the safety considerations for the Purex plant is the question of flammability limits for the diluent-air system. The present note attempts to delineate the temperature regions of concern by utilizing hydrocarbon flammability data available together with information on the vapor pressures of the systems.

An investigation of corrosion in the 327 Building's water storage basin was made to determine whether the static storage of aluminum jacketed fuel elements could cause corrosion effects that would interfere with studies of in-pile corrosion.

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.

For some time, the need has been recognized for an X-ray diffraction unit applicable to studies concerning the crystallographic characteristics of pile irradiated materials. Except for studies involving specimens of low activity, it has been impossible in the past to investigate materials which emit intense beta and gamma radiations. The exact type and degree of irradiation damage to high activity materials has been largely a matter of theoretical conjecture. The object of this report is to describe an X-ray diffraction spectrometer that can be used to collect crystallographic data from irradiated materials.

Investigation to evaluate some of the designs, techniques, and methods for fabrication of a root pass.

The purpose of this report is to show the impedance of an approximate equivalent circuit on a complex plan for various constant frequencies and different metals. The equivalent circuit represents a testing coil driven with an AC voltage and a network to represent the impedance of the metal.

Photographic observations of the behavior of uranium dioxide at high temperatures are of great value in designing and evaluating fuel elements. This paper reports the growth of uranium dioxide crystals by vapor deposition during out-of-reactor and in-reactor experiments.

A new package type facility for corrosion decontamination studies has been designed and fabricated. This report gives a description, drawings, and comments on its operation. It has been in operation since December, 1958.

Following is a description of what may be an invention in the art of compacting uranium dioxide (UO2) and other ceramic or cermet nuclear fuel materials for ultimate use as reactor fuel elements. Vibration of ceramic fuel materials at the resonant frequency of the metal encasing material, using high rates of acceleration, produces uniformly compacted high-density fuel elements.

Three proprietary high strength, low alloy steels and a high strength alloy steel were welded without a pre or post heat treatment with low hydrogen, iron powder coated electrodes of appropriate physical properties. Tensile and bend tests performed upon sample weldments indicated that they exhibited properties very similar to those for the base plate and that there were no indications of brittleness, notch sensitivity, underbead cracking, lack of fusion, hot cracking, undercutting, or crater cracking.

The chemical processing industry relies heavily on the use of rotary pumps to move massive quantities of liquids. The rotating elements of these pumps, generally of the deep-well turbine type, are submerged in the solution being pumped. This singular factor sometimes imposes a severe limitation on the choice bearings because the liquids are frequently corrosive and have poor lubricity. At the Hanford Atomic Products Operation a further complication arises from the effects of radioactivity in the solutions being transferred. Radiation and temperature can and will cause physical damage to many substances, including certain potential bearing materials such as plastics. These factors, coupled with the economics of remote operation and maintenance, have lead to the need for a test program to screen and evaluate potential bearing and journal materials.

Late in 1958 it became apparent that some fuel elements were failing in the Hanford reactors as a result of water entering through the weld. The mode of entry appeared to be first through a void in the weld, then through a non-wet area or a train of voids in the braze, and finally to the uranium core. Defective closures of a similar nature were also typical of many fuel elements which have failed in the autoclaving operation as shown in Figure 1.

Prevailing dimensional tolerance for sintered UO₂ reactor fuel components to be enclosed in metal jackets are expensively small. For instance, the PWR fuel pellet, as of November, 1955, consisted of a centerless ground UO₂ cylinder 0.3560" (+0.0005", -0.000") in diameter and 0.3527" (±0.0008") long. This pellet was designed to fit in a Zircaloy tube 0.3585" (+0.0005", -0.0000") inside diameter. The use of UO₂ shapes as pressed, or extruded, and sintered with diameter variations controlled to a tolerance of ±0.003" to ±0.005" would represent a fabrication cost saving worth considering. It is reasonable to assume that the annulus between a sintered UO₂ shape and its container will vary as a function of time of irradiation. Shrinkage from the can walls occurs with relatively low density (i.e., 85% of theoretical) sintered irradiation (cf. MTR Test GEH-4-3C). Contrary to this effect, the thermal expansion coefficient of UO₂ is about twice that of Zircaloy. Of possibly greater significance is the inevitable fractuce of dense UO₂ due to thermal stress. Such cracking effectively relocates the annular heat transfer gap to the interior of the oxide where the higher temperatures enhance the heat transfer coefficient. Thus, possibly the worst condition, with respect to heat transfer, is …