Studies were made on the safe handling of HHO/sub 3/-HF solutions in 304 L and 309SCb stainless-steel equipment under carefully controlled conditions. The corrosion behavior of both wrought and welded 304L and 309SCb was investigated in various HNO/sub 3/--HF solutions, ranging in HNO/sub 3/ concentration from 0 to 10.0 M and HF concentration from 0.01 to 1.5 M, at temperatures from 24 deg C to the boiling point. (auth)

A technical and economic evaluation is presented of a direct-cycle light- water boiling reactor designed for natural circulation and internal steam-water separation. The reference lOO-Mw(e) reactor power plant design evolved from the study should have the best chance (compared to similar plants) of approaching the 8 to 9 mill/kwh total power-cost level. (W.D.M.)

Report issued by the Argonne national Laboratory discussing a technical and economic evaluation of a direct-cycle light-water boiling reactor designed for natural circulation and internal steam-water separation. The reference 100-Mw(e) reactor power plant design evolved from the study should have the best chance (compared to similar plants) of approaching the 8 to 9 mill/kwh total power-cost level. This report includes tables, and illustrations.

A coulometric titration method was developed for the determination of uranium in HNO/sub 3/ solutions. Uranium was reduced to the (IV) state and was titrated with electrolytically generated cerium(IV). The colorimetric end point was detected automatically by a photometric technique. Interference from nitrate was eliminated by the addition of urea to the titration medium. The coefficient of variation for the analysis of uranium was 3.2% for 1-mg samples and 0.3% for 100-mg samples. With the addition of urea as much as 4.5 milliequivalents of nitric acid in a 1- to 3-ml sample was tolerated. (auth)

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

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Monthly newsletter of the High Plains Underground Water Conservation District No. 1, discussing the field of underground water. Topics include profiles of water conservation research, annual pre-plant soil moisture survey data, annual Winter Water Level measurement data, and information about the latest water conservation tips.

One of the three major categories of HAPO fuel element failures is the side corrosion type rupture. The majority of side-corrosion failures has been characterized by oval or tear-drop shaped flow patterns containing evidence of accelerated corrosion. Thorough examination of many of these so-called `hot spot` failures has indicated the failure was caused by poor heat transfer associated with misalignment, dimensional distortion or poor jacket-to-core bonding. It has been postulated that misalignment of the fuel element is a necessary condition for formation of hot spots under the present reactor operating conditions. Neither tru-line contours nor X-8001 alloy are successful in the prevention of misalignment and associated ruptures; therefore, it has been proposed to test the effectiveness of projections on the side of the fuel element toward preventing fuel misalignment in ribbed process tubes. A previous test of this element termed the `bumper fuel element` was encouraging; however, it failed to provide the conclusive proof required to justify a large-scale demonstration loading. Supplement A to the basic test was written to obtain necessary preliminary data. This report presents an outline of further testing required to accelerate evaluation of the bumper concept.

The objective of this test is to determine the DR-Reactor effluent systems characteristics under 95 degrees Celsius bulk temperature operation. This proposed bulk temperature increase from 93.5 to 95 degrees represents a 33% decrease in the bulk temperature suppression below the boiling point. A major aim of this test will be to evaluate the degree of increased maintenance at this higher temperature operation. The basis and justification, test preparation and instrumentation, procedure, costs, outage time, hazards, standards, and responsibilities are discussed in this document.

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Irradiations were made on small specimens of U-5 wt.% Zr-1.5 wt.% Nb alloy with a wide variety of fabrication histories and heat treatments in order to determine the optimum heat treatment for the fuel plates for the Experimental Boiling Water Reactor (EBW). In the time available, a heat treatment could not be found which simultaneously conferred dimensional stability and corrosion resistance to the alloy. The most effective heat treatment for dimensionally stabilizing swaged or round-rolled material was a 24-hr isothermal transformation from the gamma phase at 650 deg C. This heat treatment was subsequently used as a basis for the heat-treatment specifications for the EBWR fuel plates. In later studies on specimens cut from plates it was learned that the alloy could be adequately stabilized against irradiation growth and also made corrosion- resistent by first reducing the plate 12% in thickness by cold rolling followed by a 24-hr isothermal transformation from the gamma phase at 665 deg C, and finally quenching from 800 deg C. Irradiation growth rates of plate specimens were effectively reduced by the presence of metallurgically bonded Zircaloy-2 cladding. Flat-rolled material under irradiation generally increased in length and width and decreased in thickness. (auth)

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Abstract: "Available data on wind persistence and wind forecasting capability have been applied in estimating the probability of a fallout pattern shifting from an uninhabited safe sector into a populated region. Safety probability is computed from win variability, forecasting accuracy, initial height and particle size of radioactivity landing at a point in the predicted fallout pattern, predicted wind speed, length of forecast period, and safe-sector angular width."

Supersedes ITR-1421. Twelve large-diameter buried conduit sections of various shapes were tested in the 60- to l49-psi overpressure region of Burst Priscilla to make an empirical determination of the degree of personnel protection afforded by commercially available steel and concrete conduits at depths of burial of 5, 7.5, and 10 feet below grade. Essentially, it was desired to assure that Repartment of Defense Class I, 100psi and comparable radiations, and Class II, 50-psi and comparable radiations, protection is afforded by use of such conduits of various configurations. Measurements were made of free-field overpressure at the ground surface above the structure; pressure inside the structures; acceleration of each structure; deflection of each structure; dust inside each structure; fragmentary missiles inside the concrete structures; and gamma and neutron radiation dose inside each structure. All buried conduit sections tested provided adequate Class I protection for the conditions under which the conduits were tested. Standard 8-foot concrete sewer pipe withstood 126-psi overpressure without significant damage, minor tension cracks observed; standard 10-gage corrugated-steel 8-foot circular conduit sections withstood 126- psi overpressure without significant damage; and standard 10-gage corrugated- steel cattle-pass conduits withstood 149-psi overpressure without significant damage. Durations of positive pressure were from 206 to …

It has been thoroughly documented in previous Bridgeport Brass Company reports that a problem exists in the production of HAPO I and E ``O`` size slugs via the extrusion route due to an excessive number of rejects for I.D. seams. As a result of a previous short-run experiments, it was suggested that smaller I.D. billets be employed, thus reducing the reduction ratio on the I.D. It was thought that this factor may contribute to the seam formation in the extreme flow demanded of the metal. It was also thought that the long billets, half of an ingot, could contribute to this defect. Therefore in this extrusion, some short billets, one-third of an ingot, were employed. Other methods or techniques for reducing the incidence of this I.D. defect will be studied later. It was intended to limit the present experiment to a study of the potential reduction involved in the above; namely, billet length and billet I.D., in an extrusion experiment with sufficiently large quantity of billets to give a fairly large number of slugs. By proper design, other contributors to the defect were held constant or randomized. This experiment proved definitely that a reduction in I.D. rejects due to seams …

Two fuel elements were removed from the Experimental Boiling Water Reactor and examined in a hot cell. The elements had maximum burn-ups of 0.11 and 0.39 at.%. Both were disassembled and sampled for the evaluation of the effects of in-pile operation and radiation damage to the fuel. The fuel elements were in gcod condition with no ruptured.cladding, core-clad nonbonds, or excessive fuel-plate swelling or warpage. Thin samples cut from the fuel plates in element ET-51 warped and cracked, suggesting a relieving of locked-in stresses and indicating that after 0.39 at.% burn-up the fuel cores were hard, brittle, and highly stressed. The rate of fuel-plate volume increase owing to the burn-up of uranium was 6 to 7% DELTA V per at.% burn-up. Hydrogen was picked up by the fuel plates under reactor operating conditions with the probable forraation of isolated areas of small announts of zirconiura hydride. Annealing studies on sections of fuel plate at 500 and 550 deg C indicated bulk volume increases of 1 to 2% and 5 to 10%, respectively, after 500 hr. A 600 deg C anneal resulted in a bulk volume increase of 17% after 45 hr. (auth)

This bibliography is selective, unclassified, and covers the period 1950 - April 1960. Sources consulted were the following: Applied Science and Technology Index (Industrial Arts Index), Chemical Abstracts, Engineering Index, U.S. Atomic Energy Commission Nuclear Science Abstract, Armed Services Technical Information Agency Title Announcement Bulletin, and Publisher's Catalogs.

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Use of X-8001 Al alloy as cladding for Hanford reactors was initiated because of superior (laboratory) resistance to intergranular corrosion over that of C-64 alloy. However, since severe pitting attack was observed intermittently, an evaluation was carried out on X-8001 alloy fuel element cladding.

This specification covers the design, fabrication, testing and furnishing of two fire pumps complete with drivers and appurtenances.

Wood, coke, and coconut-shell charcoals were evaluated for fluorine entrapment. The coconut-shell charcoal produced the smallest amount of solid and liquid reaction products. Efficient removal of fluorine was accomplished by the coconut-shell charcoal in a 5-in.-diameter reactor with a feed containing 25% fluorine at flow rates from 100 to 400 scfh and reactor-wall temperatures of 1200 to 1800 deg F. (C.J.G.)

Report that "describes an instrument that nondestructively tests uranium fuel element cores by ultrasonic methods" (p. 2).

This document details activities of the Fuels Preparation Department during the month of June 1960. (FI)

The problem of radiolytic gas diffusion into a bubble of fixed radius is solved. A constant source of radiolytic gas is assumed. The concentration of gas at the bubble surface is related to the pressure within the bubble by Henry's constant. (W. L.H.)