Isotopic assays have been made on the Cr in samples from 14 different chrominiferous minerals from different geographic and meteoritic sources. The results of the assays indicate that it is not possible to state unequivocally that variations in isotopic composition have been observed. (auth)

Measurements of thorium resonance-capture activations relative to those of Au/sup 197/, which were made in the Peach Bottom (HTGR) critical assembly, are analyzed and compared with thorium resonance activations calculated from tabulated values of the resonance parameters. The new vanadium-subtraction method that was applied is shown to measure thorium resonance captures relative to those of Au/sup 197/ to within several per cent in this geometry. The measurements show that the lowest several resonances of thorium capture less strongly than would be predicted by the listed parameters. This reduced capture is such that the thin-limit, epicadmium 1/E-spectrum resonance integral is about 84 b instead of the 96 b predicted by listed parameters; the thorium resonance captures for the critical assembly are about 6% less. The approximations currently used in calculating resonance captures in annular lattices probably underestimate captures by several per cent. The 300 deg K resonance-capture rate of ThO/sub 2/ in the annulus is indicated to be several per cent larger than that of metallic thorium, presumably because of a solid-state effect. 46 references. (auth)

The status of aqueous processing methods for thorium fuels is summarized, with principal emphasis on the stainless steel-clad ThO/sub 2/UO/sub 2/ type. Data were obtained principally from laboratory-scale experiments with fully irradiated fuel samples and engineering-scale tests with unirradiated fuel. Stainless steel cladding was easily dissolved with 4 to 6M H/sub 2/SO/sub 4/ (Sulfex process) or 5M HNO/sub 3/-2M HCl (Darex process) in LCNA (Nionel type) or titanium equipment, respectively, in semicontinuous or batch equipment. Uranium losses to the decladding solutions were approximates 0.3% and 3 to 5% for the Sulfex and Darex processes, respectively, with fuel irradiated to approximates 20,000 Mwd/ton of core. The uranium was readily recovered from the Darex decladding solution in the acid Thorex extraction process. The ThO/sub 2/UO/sub 2/ core was dissolved in 13M HNO/sub 3/ -0.04M NaF-0.1M Al(NO/sub 3/)sub 3/. Uranium and thorium can be recovered from graphite-base fuels by disintegration and leaching with 90% HNO/sub 3/, grinding and leaching with 70% HNO3, or combustion followed by dissolution in fluoridecatlyzed nitric acid. Uranium and thorium were recovered from nitric acid solutions and separated from fission products by extraction with 30% tributyl phosphate in Amsco in the acid Thorex process. The use of an acid …

Research and development on metallurglcal aspects of pressurized-water systems is summarized. A survey was made of the methods of determining fuel burnup. The mechanisms and kinetics of the loss of boron during heating at 1135 deg C in various dynamic environments were determined. A model was developed to quantitatively characterize the UO/sup 2/ dispersion microstructure of roll-clad fuel plates relative to an ideal'' dispersion. In order to avoid the loss of boron from UO//sub 2/- stainless steel dispersion fuel plates during fabrication, studies were carried out on a refractory glass containing 4 wt.% B/sub 2/O/sub 3/. By using lowsilicon elemental powder, the undesirable reaction between Eu/sub 2/O/ sub 3/ and Si was eliminated; and 13 full-size SM-1 absorbers were fabricated. Work was continued on the borongradient neutron absorber concept. A design was studied for preparing a composite control rod having an upper section made of a boron-gradient dispersion and the lower tip made of Eu/sub 2/O/sub 3/ and stainless steel. Two fuel elements were examined after significant exposure in SM-1. The examination of the miniature boron-iron samples in the final phase of the MTR irradlation test was performed. Twelve miniature test specimens containing 20, 30, or 40 wt % Eu/sub …

Techniques for forming polycrystalline LiF from the melt and for fabricating test specimens were developed and evaluated using single-crystal LiF as a control. Large -grain polycrystalline specimens tested in fourpoint loading always showed some plastic deformation (0.078 to 0.798%) before fracture, but the plastic flow was sharpiy reduced from that of single crystals. An etch was developed revealing dislocations on all crystallographic faces of LiF. Details of plastic deformation in polycrystalline material were investigated. Deformation was inhomogeneous among the grains of an aggregate because of differences in orientation with respect to the applied stress, also within individual grains because of interactions between adjoining grains. Grain boundaries were barriers to slip, but stresses resulting from slip in one grain were transmitted to neighboring grains and often caused local deformation near the boundary. Because of local stresses, local slip systems operated although the resolved shear stresses on them from the applied load were below the critical yield stress. In one case, slip occurred on an (010) plane. Three-grain junctions were areas of high residual stress. Fractures originated at boundaries at or near 3-grain junctions, not as a result of inherent boundary weakness but rather because of high stresses developed at the boundary. …

Based on process development work at ORNL, a facility was designed for demonstration of the pot calcination process with a variety of high-level radioactive wastes. In this facility, operational and control problems associated with an integrated process can be identified and solved, procedures to improve characteristics of the calcine can be studied, and important aspects of calcine temperature distribution and fission product behavior can be observed. Installation of the facility is planned for the Hanford Atomic Products Operation. The facility can be installed in any cell or cells having certain basic features. (auth)

Aqueous methods for recovering uranium from BeO- and Al/sub 2/O/sub 3/- base gas-cooled-reactor fuel elements are being evaluated. Two methods for processing Hastelloy-X--clad pelletized BeO-base fuels containing 60 to 70% UO/ sub 2/, such as the GCRE and MGCR, seem feasible. One method involves mechanical stripping or chopping of the cladding followed by leaching of the uranium from the fuel pellets with boiling 6-l3M HNO/sub 3/. In the other method the cladding and UO/sub 2/ are dissolved in boiling 2M HNO/sub 3/-4M HCl. In either case, most of the BeO matrix remains as an undissolved residue. Pellets containing 70% UO/sub 3/ dissolved completely in less than 20 hr in boiling 8M HNO/sub 3/ containing either 2M H/sub 2/SO/sub 4/ or 0.5M HF, producing solutions containing 4 g of uranium per liter. Fuels of high BeO content, e.g. BeO--5% UO/sub 2/, dissolved only slowly in boiling aqueous reagents. Highest initial rates were in sulfuric acid solutions, log (Rate, mg min/sup -1/cm/sup -2/) = 0.223 (H/sub 2/SO/ sub 4, M) - 2.8l and in HF--NH/sub 4/F solutions. ln boiling 5-8M NH/sub 4/F the initial dissolution rate increased from 0.07 to 3.5 mg min/sup -1/cm/sup -2/ as the HF concentration increased from 0 …

By equating net radiation to space to I/sup 2/R heating in the SNAP-2 instrument compartments, and neglecting internal radiation between compartments, the panel surface equilibrium temperatures were computed for hot and cold temperature extremes. These extremes are defined by: (1) cold - pre-start phase in the shade, (2) hot - full power phase facing the sun. The results of the heat balances for hottest and coldest cases are presented graphically. These figures relate the equilibrium surface temperatures to the I/sup 2/R heat load dissipated by the panel surface for hot and cold orbits before and after startup. Included as parameters are effective panel area (dependent on Agena-interface design details) and the ..cap alpha../epsilon ratio for the surface coating. If ..cap alpha../epsilon = .3/.9 and A = 1 ft/sup 2/, the extremes of temperatures suffered are from -191/sup 0/F to +212/sup 0/F. This study shows that the normal R/C surface coating (..cap alpha../epsilon = .3/.9) is acceptable regarding allowable maximum surface temperatures, if the effective panel area is not less than 1 ft/sup 2/. It also indicates that further work is warranted regarding internal radiative distribution of heat in order to limit the lowest temperatures to -65/sup 0/F.

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This is the monthly report for the Hanford Laboratories Operation, November 1962. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This is the monthly report for the Hanford Laboratories Operation August 1962. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

A hazards summary report for the planned installation and operation of an L-77 Laboratory Reactor of the University of Nevada is presented. Site data, including information on the geography, geology, seismology, climatology, and hydrology of the area in which the reactor will be installed are included. The reactor site and administiation of the reactor facility are described along with the reactor, its uses, and its performance characteristics. Analyses of the nuclear, radiation, and operational hazards are also included. (auth)

ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiated …

An in-pile pump loop, designed to fit within horizontal beam hole HB-2 of the Low-Intensity Test Reactor (LlTR), was used to circulate an aqueous thoria- urania slurry while exposed to reactor irradiation. The total loop volume was about 1600 ml, including pump and pressurizer, but the slurry was confined to the 900-ml volume of the main loop stream by means of a sintered stainless steel filter. The filter was an important feature of the loop design in that it provided a thoria-free filtrate as a purge stream to the pressurizer and pump bearings to prevent entry and accumulation of thoria in these two regions. Corrosion-test specimens of Zircaloy-2, titanium, and type 347 stainless steel were placed in the loop at three different locations for exposure to three different levels of irradiation. Duplicate sets of specimens in each position were exposed to flow velocities of 8 and 22 fps, respectively. For the in-pile irradiation, thorium oxide containing 0.43 wt of enriched U, based on Th, was used. This thoria-urania was produced by air calcination at l225 deg C of coprecipit.ited oxalates and had a me.in particle size of l.7 mu . A Pd catalyst w-as dispersed in the slurry for liquid-phase …

During the course of the Christmas Island portion of Operation Dominic, a fairly simple, inexpensive method of making a rapid fireball yield determination was developed. In order to check the zero time alignment of the skysweeper antiaircraft mount on the burst point at zero time as provided by the Sandia 584 radar system, a pinhole camera camera was installed on the mount. This was helpful in determining whether or not the photographic and photoelectric diagnostic systems were observing the appropriate volume of air excited by prompt gamma and neutron radiation. After viewing the initial circular image on the Nambe event, it was decided that an attempt at a prompt fireball yield number would be made. Ad hoc scaling laws were used for the duration of the operation and the instant fireball yield number was included in several TWXs of early shot data. This report provides calculations, graphs, and illustrations which show that with the data now in hand, there is a simple method of determining a yield number to generally better than 10 percent. This number can be available within minutes after the detonation.

This document details activities of the Irradiation Processing Department during the month of August, 1962. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; and NPR Project.

This document details activities of the Irradiation Processing Department during the month of November, 1962. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

The interests of several departments at Hanford are involved in the planning, execution and evaluation of the results of the KER loop testing effort in support of the NPR fuel program. The varied interests and activities of the participating groups must be well-integrated if effective use of our limited testing capability is to be made. The purpose of this report is to help achieve this integration by summarizing the current thinking on the goals of the NPR fuel testing program and by presenting the current loop schedule.

Separate abstracts have been prepared for four sections of this report. The sections deal with reactor engineering and physics, nuclear physics, instrument development, and applied mathematics. (N.W.R.)

Six procedures, in stepwise form, that can be used for the determination of plutonium in several forms by controlled-potential coulometric titration are presented. (auth)

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As TORY II-C is presently designed, a nuclear excursion would occur if the core were flooded with water. This is true even if all of the existing control rods were fully inserted. Indeed ANGIE calculations indicate that the reactor would be about 30% super-critical in such a case. There are several methods by which the core may be forced sub- critical under these extreme conditions. We will here consider only the method of introducing, directly into the core, a material which strongly absorbs neutrons. It must be possible to remove this excess `poison` prior to, or during the boost phase. Since the computer codes can be trusted to only approximately 3%, we will, for safety, insist on 40% negative reactivity to be introduced by the excess poison.

During the normal operation of reactors ionic, particulate and colloidal substances in the coolant are activated as they pass through the neutron flux. The parent substances for the activated materials come from several sources: transient corrosion products from the coolant system piping and fuel element cladding, coolant additives and coolant impurities. Previous studies have shown that a major portion of these substances do not simply pass through the neutron flux field but undergo complex sorption phenomena that result in effluent radio-nuclide levels considerably in excess of what would be expected from activation on a once-through basis. This report is being written to outline studies that will provide activity inventory data that will aid in defining radionuclide retention reactions in reactor coolant systems.

The objective of this production test is to evaluate the effect of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining of coextruded Zr-2 jacketed rod samples.