This report presents a study of the stresses induced in the elbow and downcomer of the B and C Reactors (bulk effluent systems) by flow momentum and pressurization effects. It is desirable to ascertain the limitations on the bulk outlet temperatures and flow rates from a stress viewpoint; the elbow, top coverplate of downcomer, and top baffle plate being the most severely stressed components.

Brittle bonds and the factors controlling their formation have been of substantial concern in the production of dip canned fuel elements. Detection of brittle bonds has been by the chisel test and by metallographic examinations. At best, these are qualitative tests and do not establish the degree of brittleness. For this reason bond tensile strength analysis has been suggested. Tests have been run to determine if a change in canning variables could be detected by a change in bond strength.

This report is a summary of projects worked on in support of the production reactors at Hanford. The projects include criticality studies, from tasks associated with fuel element reprocessing to shipments of slightly enriched uranium. They include studies of neutron cross sections for different reactions and neutron flux measurements in different reactor locations, as well as design studies for future reactor projects.

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The masonite in the old pile shields is deteriorating because of the temperature conditions to which it is subjected. One method of mitigating this deterioration is to reduce neutron leakage to the shields by utilizing the fringe tubes for target material. One possible target material is depleted uranium, i.e., uranium from which the bulk of the isotope 235 has been removed. The purpose of this test supplement has been to irradiate a fringe depleted pattern to establish the enrichment loading of E metal necessary to maintain reactor production, to determine the depression in shield temperature which could be obtained by such a loading, and to establish the operational economics of this method of production. The latter of these goals is to be the subject of a later report. With full length depleted uranium charges in the outermost fringe process tubes the heat generation rate in the adjacent shields is reduced to approximately 75 per cent of the value for a natural uranium loading under comparable reactor conditions. The reactivity compensation can be achieved by the use of one and one-half full length E columns for each depleted column where the E charges are placed in the second an third tube …

Assistance has been requested by Financial Operation to determine the percent of firm and interruptible power used in each area for cost distribution purposes. Also, Power Operations have suggested that the power cost distribution be reviewed. In consideration of these requests, and with the use of improved demand instrumentation within the 151-B, D, and F substations, a proposal for equitable IPD electrical power cost distribution between areas is presented for acceptance or comment by Area Management.

On the basis of numerous rough calculations it has generally been assumed that if the water flow to a tube were to stops while pile operation continued. The tube central temperature would increase in a matter of seconds to dangerously high values. In several seconds the water around the central slugs would flash to steam, shortly thereafter the aluminum cans would melt and this would give good contact between the slug and the tube which would quickly melt. The bare uranium would react with the steam and possibly the pile gas atmosphere. Radioactive Xenon and fission products would be spread through the gas system, and the molten aluminum and uranium would fill up all of the cracks in the graphite and the tube would be impossible to discharge by all normal methods after the pile is shut down. If the pile were operated with this tube blocked off there would still be the problem of exceedingly high graphite temperatures around it, and the spread of contamination in the gas system. Because these problems are expected with an undetected water failure in a pile, where the operation is maintained, a program is underway to ensure proper and sufficient tube instrumentation to …

The objective of this test is to authorize large-scale irradiations of X-8001 (formerly designated M-388) alloy jacketed fuel elements in order to evaluate their suitability to the reactor process.

The author tells about some of the most interesting aspects of recent research on the synthetic transplutonium elements. The amount of recent information on these elements is obviously too much to cover completely in the time that has been placed at my disposal. Therefore, in planning my talk, I have attempted to choose those topics which have the broadest implications for the whole transuranium field of research. Although much important and interesting research is, of course, being done in many laboratories, I have chosen examples mainly from the work in our own laboratory, the Radiation Laboratory at the University of California in Berkeley. I shall cover this information about the known transplutonium elements, listed in Slide No. 1, by discussing them in order of increasing atomic number, and I shall conclude with some thoughts concerning the prospects for elements with higher atomic numbers than any that have been produced and identified up to the present time. For purposes of orientation, Slide No. 2 shows the position in the periodic table of the presently known and the future transuranium elements. The transplutonium elements through element 103, together with the five preceding elements, are members of the 'heavy rare earth', or actinide …

Two eddy-current electromagnets to act as brskes were designed and installed in the sodium-cooled nuclear pcwer reactor SRE to throttle sodium flow throughout the reactor after shutdown in order to maintain a constant reactor temperature gradient. One brake was used on the primary piping system, the other on the secondary system. It was determined that the eddy-current brake should cause a dragging pressure of 0.3 psi at a flow rate of 12 gal/min. The flux density necessary to produce this pressure was calculated, and the coil ampere- turns required to produce this fiux density were determined. Both brakes were controlled by thermocouples and performance was satisfactory. (M.C.G.)

A 1585 hr run was completed on the 400A-1 pump with a 0.4 m UC/sub 2/SO/ sub 4/ fuel solution as the pumped fluid, during which it was attempted to operate with gas bubbles in the solution. The gas content ranged from 0.56 ml/gm of solution to 1.189 mi/gm of solution and the gas other than CO/sub 2/, which was assumed to he nitrogen, varied from 0.39 ml/gm of solution to 0.966 ml/gm of solution. Because of uncertainties in the CO/sub 2/ solubility it cannot be stated conclusively that the liquid was or was not saturated. However, the loss of metal at a number of places due to corrosion and/or gas bubble erosion makes it advisable not to operate a pump of this exact design at or near gas saturation. The 1585 hr above are the last of 13185 hrs maintenance free operation. A general examination of the pump dlsclosed severe corrosion and softening of the Stellite 12 overlay on the 347 stainless steel thrust bearing shoes. The thermal barrier seal weld had cracked and there was noticeable corrosion in the region betwoeen the thermaI barrier gasket and metallic O'' ring seal. (auth)

The laboratory and pilot plant evaluation of Stanrock concentrate is presented. This concentrate meets all FMPC impurity tolerance limits except that for thorium content. The thorium limit can be met by the addition of phosphate as required for thorium-containing uranium concentrates. Laboratory digestion and extraction tests indicate that this material should present no problems in pulse column processing. A pilot plant pulse column test (a test utilizing a modified two-inch-pulse-column system) indicated that this material should present no operational problems in the National Lead Company of Ohio Refinery. (auth)

A stress analysis is presented of a spherical shell subjected to a line load acting along a latitude circle in a direction normal to the plane of that circle. (auth)

Document distributes the charted information given at the briefing

S>Of primary interest in a military reactor is the desirability of quick and easy maintenance of as much of the system as possibie throughout the lifetime of the power plant. It is therefore desirable that no serious amounts of long- lived activity build up in equipment outside the core. An investigation was made to determine the activity due to small amounts of cobalt and tantalum in the stainless steel core of the APPR-1. The investigation indicates that equal weight percentages of Ta and Co would build up about equal activities in the core after 1.5 years of operation. The use of 347 stainless steel containing a nominal 0.2 wt.% Co would therefore produce a more serious activation problem than is now being experienced wlth the present 304L core. A desire to reduce the production of long-lived activities would then require specifying a low tantalum as well as a low cobalt content if 347 stainless steel were used for the second APPR-1 core. Fabrication of a low cobalt 304L core is therefore an attempt to reduce and to simplify the activation problemn (A.C.)

The reaction between uranium dioxide and aluminum has been studied at 600 deg C and below. At 600 deg C the visible reaction occurred within a few hours, while, for example, at 500 deg C and below, the specimens had to be at temperature for several days in order to show signs of any reaction. The method of manufacture of uranium dioxide had a pronounced effect on its compatibility. Granulated and high-fired oxides showed lowest, and the steam-treated oxide showed maximum, reactivity. The fact that finer oxide powder partlcles react faster than coarse ones was according to expectations. Mallinckrodt U/sub 3/O/ sub 8/ powder of very fine particle size reacted severely with aluminum during a 1 1/2-hr test at 610 deg C. (auth)

An internal flux suppressor will be used for the APPR-1 Core II instead of the external suppressors that are employed in Core I. The reference design of the integral flux suppressor for Core II is as follows: materials Eu/sub 2/O/sub 3/ in stainless steel matrix; compositions 23.2 wt.% Eu/sub 2/O/sub 3/; and dimensions, 0.02 in. x 2.281 in. x 0.875 in. The survey on the cffectiveness of this referencc design has shown that thc peaking of the flux in the control rod fuel element tip is eliminated throughout the lifetime and therefore eliminating the possibility of local boiling in this region. (auth)

The insoluble residues produced during the reprocessing of certain nuclear fuel elements containing aluminum, silicon, and uranium were investigated with respect to particle size, shape and distribution, composition, and surface- active tendencies. The fuel material samples studied contained from 0.4 to 7.0% silicon by weight (the high analysis represents a cast base AlSi alloy). The fuel materials were dissolved in mercury-catalyzed nitric acid. Two types of solid residue were produced by actual fuel dissolution. One was a finely divided material, brown in color and the other was black, crystalline material relatively large in size. Only the black crystalline material was obtained when cast AlSi was dissolved in nitric acid. Spectrographic analysis of the residues showed that silicon and alumimum were the major constituents. X-ray diffraction analysis of each type indicated the presence of elemental silicon only. The x- ray pattern obtained with the brown material showed lower intensities indicating the presence of amorphous material. Under a magnification of 970 diametems, the crystals observed appeared to be non-uniform, shallow platelets of irregular shape. The bulk of the residue is remmed by a 10-micron filter: however, the resulting filtration rates are very low. (2.0 gallons per hour per square foot). Sedimentation data …

Critical-assembly studies were made to provide engineering and physics data to aid in developing the Gas Cooled Reactor Experiment-1 (GCRE-l). Measurements of critical mass, flux and power distributions, and shutdown worth of the GCRE-1 mock-up safety and control blades were obtained. The critical assembly consists of aluminum tubes containing four concentric stainless steel cylinders wrapped wtth highly enriched uranium foil. These tubes are supported at each end by grid plates aiid arranged to approximate a right circular cylinder. The entire core structure is supported within a tank which can be filled remotely with moderator water. A void beneath the core structure and the air above it represent the gas plenums of the GCRE-1 reactor. Critical mass and flux power distributions were determined for several cases of four basic cores. The thermal utilization, measured in two core configurations, was 0.780 for the reactor wtthout burnable poison or a lead reflector. The temperature coefficient of reactivity, measured in two cores, was positive at room temperature but negative in the proposed operatingtemperature range. The total reactivity effect in going from 20 to 80 deg C was a positive 0.22% DELTA k/k. The worth of control and safety-shutdown blades of several compositions and sizes …

An order of magnitude estimate was made a determine the cost of producing oxygen-18 enriched water by the equilibrium distillation of water. Three isotopic purities and two production rates were considered. Costs varied from per gram for 3% oxygen-18 enriched water produced at a rate of 100 grams per day to 5 per gram for 99% oxygen-18 water produced at a rate of one gram per day. (auth)

At the request of the manager of Process Engineering, a study was made of the fuel element recovery operation and its relationship to the HAPO uranium scrap losses. Three fundamental questions formed the basis for this study. They are: What is the amount of the yearly uranium scrap loss at HAPO? What contribution does the existing fuel element recovery process make to this loss? What should be done to reduce this loss? The purpose of this report is to answer these questions and make specific recommendations aimed at reducing these scrap losses. The basis for these recommendations will be explained in the DISCUSSION section of this report.

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The continuous calciners at UO{sub 3} Plant are of a new design which was developed at HAPO and placed in service late in 1956. The heat transfer troughs are considered to be the most vulnerable parts of the calciners because of their high operating temperatures. Thermal stresses are calculated to be quite high, and when added to the direct mechanical stresses from powder load etc., come close to the limiting safe stress for stainless steel at the operating temperature. It is felt that trough failure will be progressive as a result of creep type distortion and cracks at stress concentration points. The higher the stress (and temperature), the more rapid the rate of failure will be. All possible steps have been taken to limit maximum trough temperature, rate of change, and variations from one part of the trough to another. These steps were not always successful and various troughs have been subjected to rather severe temperature shocks as well as high mechanical stresses due to agitator failures. Despite these difficulties, no signs of failure could be detected by visual inspection. It was decided, therefore, that a more complete examination should be made. This examination was made to determine the present …

A general non-uniform burnup program was developed to determine the lifetime of the APPR-1. The calculation is performed using two one dimensional multi-region burnout calculations. The approach to the problem, the equations, and derivation of burnout equations are presented. The results are plotted and compared with the rod bank position as measured at Fort Belvoir. On the basis of these calculations the expected total energy release of the APR-1 is 13 Mw-yr. (auth)