ABS> The removal rates of iron, nickel, and chromium from synthetic stainless steel waste solutions during electrolysis over a mercury cathode were studied. The loading capacity of the mercury for the stainless steel metals was estimated on the basis of laboratory experiments to be about two% by weight. The laboratory data indicated that, at an electrode potential of --1.80 voits vs S.C.E., 85 ampere-hours per liter of waste removed essentially all of the stainless steel ions from a sulfuric acid solution containing 0.13M metal ions at 35 deg C. (auth)

Progress is reported on thermonuclear research. Separate abstracts were prepared for 8 of the 10 sections. Design and engineering service reports and notes are given in the remaining sections. (M.C.G.)

Using the group vibratibn method of McMurry, the normal frequencies and coordinates of propane and three of its symmetrical deuterium substituted compounds were obtained. The force constants used were taken from a variety of previous works on hydrocarbons. The results give reasonable agreements with the experimental frequency and mode assignments of others. (auth)

A test was conducted on August 12 to 13, l961, on the operation of the four reactor coolant loop relief valves and the four purification system relief valves. The results indicated proper operation. (D.L.C.)

The multiple hydroclone system removed l83 grams of corrosion product solids in l814 hours of operation during reactor run 2l. The low removal rate was attributed to plugging of multiclone feed ports that presumably occurred during the latter part of run 20. After modifications to the reactor core and removal of the multiclone unit at the end of run 21, the reactor core was backflushed with the flow direction in the core loop reversed. During this period, the single hydroclone removed 205 grams of solids in 10.5 hours of operation. (auth)

A discussion is presented concerning the preliminary design and hazards evaluation of a H-cooled in-pile experimental loop for operation in the large double-walled through-tube in the Experimental Gas-Cooled Reactor (EGCR) at Oak Ridge. This loop is designed to permit experimentation with full-scale fuel element configurations up to 8 in. OD, at inlet gas temperatures of 600 to 950 deg F at 300 psig, and experimental power levels up to 500 kw. The results of a preliminary hazards evaluation indicate that a loop of this type can be safely operated in the EGCR. The H flammability hazard is controlled by blanketing all H-filled pipes and components with a sufficient quantity of nonreactive gas, such as He or CO/ sup 2/, to produce a noncombustible mixture for all credible H- release situations. (auth)

Since brazing was the method selected for joining the stainless steel SM- l reactor fuel element, corrosion studies were conducted on various potential brazing alloys to evaluate their resistance under the approximate pressurized- water conditions of the SM-1. The program consisted mainly of testing type 304L stainless steel T'' joints brazed with selected alloys in quiescent, degassed, and deionized autoclaved water at 565 deg F under 1200-psi pressure. In the initial phase of the investigation, tests were limited in duration to l000 hr in order to quickly screen some 18 potential alloys for longer time testing. Based on weight-change data and the metallographic examinations, five of the 18 alloys exhibited sufficient corrosion resistance to warrant further investigation. These alloys were subjected to autoclave tests of 12 and 16 months. In these extended tests, 1 cc O/sub 2/liter and a mixture of 1 cc O/sub 2/liter plus 50 cc H/sub 2/liter, respectively, were added to the water to more closely simulate SM- 1 reactor water conditions and to evaluate the effect of different gaseous additions on the corrosion behavior of the alloys. On the basis of weight-change data and metallographic examination after long-term exposure of the tested stainless steel-base joint; these …

In order to study resonances in the {Sigma}-{pi} system, we have analyzed reactions in which a {Sigma} hyperon and two or three pions are produced in K{sup -}-p interactions at 1.22 {+-} 0.040 and 1.51 {+-} 0.050 GeV/c incident K{sup -} momentum (i. e., 1895 and 2025 MeV center-of-mass energy), using the Lawrence Radiation Laboratory's 72-in. hydrogen bubble chamber.

A procedure is described for the sequential separation and determination of Arsenic-76 and Phosphorous-32 in reactor effluent water. The analysis can be performed in less than one hour and the arsenic fraction is scanned immediately after separation on a multi-channel gamma spectrometer. The discrete gamma energy of .56 MEV for Arsenic-76 makes further purification unnecessary. The organic phase containing the Phosphorous-32 is dry mounted on a one-inch stainless steel dish for beta counting.

The feasibility of operation with an enriched core to ``override`` the xenon transient and thus reduce minimum downtime has been reasonably well established. The principle economic disadvantage of the present technique is associated with the use of boron splines for compensating the reactivity effect of the core during normal operation. The boron-containing spline parasitically absorb the excess neutrons produced in the core thus wasting these surplus neutrons during the operation when the enriched core is compensated by poison; also, because of the relatively slow removal rate of irradiated splines, a significant portion of the potential downtime reduction is used to remove the splines from the core. The purpose of this report is to point out incentives and alternate methods for improved poison control with the xenon override core and to present current plans for resolving these limitations to the xenon override concept.

The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

A production fuel element which failed in tube 1372-H on March 12, 1962 at 263 MWD/T exposure was transferred to the Radiometallurgy Laboratory for examination to determine the cause of failure. The fuel geometry was 0 III N-B and the parent Lot was KZ-135 D. The failure was the result of mechanical damage to the cladding and cap at the male end prior to irradiation. A scrape or gouge at the male end reduced the cladding thickness drastically so that water entry to the uranium occurred after an incubation period of operation. Fabrication was not a factor in the failure as the male weld was found to be in good condition when metallurgically examined. The element as received may be seen in Figure 1. The mechanical damage on the cap in line with the rupture cavity was plainly visible. The formation of aluminum oxide over the marred surface indicated that the damage occurred prior to irradiation. The cladding was missing from over the rupture cavity and the fragments attached to the edge of the cap were extremely thin. It is possible that the entire cladding thickness was removed by the mechanical damage prior to or during charging. A diagonal split …

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It is noted that in high-temperature He-cooled graphite reactors, sufficiently high levels of gaseous impurities can lead to transport and corrosion effects. The possible effects of these reactions in graphite-moderated reactors designed to operate at a He-coolant pressure of a about 20 atm. were investigated. Results are included on C transport, steam-graphite reactions, and deposition of C on surfaces. (J.R.D.)

Temperature control was achieved by varying the gas mixture, and therefore the thermal conductivity, in a gas annulus surrounding the irradiation capsule. Control systems were used for over one year and maintained capsule temperature to e due to uraniu 10 l F. The system may be used equally satisfactorily with either fueled or nonfueled capsules. The reliability of the system was extremely high, and all maintenance was limited to the readily accessible instrumentation. (auth)

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A process that was developed for the preparation of particles of thorium- uranium dicarbide is described. Granulated particles composed of powdered UO/sub 2/, ThO/sub 2/, and carbon were suspended in a bed of graphite flour and heated to convert the oxides to dicarbides. The particles were then melted to densify and spheroidize them, the graphite flour keeping them separated so that they would not flow together. (auth)

Amplitude-dependent internal friction evidence is presented which demonstrates the strong differences in dislocation mobility which exist between the two classes of cubic metals: the face-centered cubic, represented by Al and Cu, and the body-centered cubic, represented by Ta, Nb, Mo, and W. Amplitude- dependent internal friction due to dislocation motion results from the hysteretic unpinning and repinning of dislocation loops from impurity pinning points when the oscillating measuring stress exceeds the breakaway stress value. Results of experiments indicate that the modulus defect measured at 4.2 nif- K in the fcc metals is considerably larger than that associated with the Bordoni peak alone, while the modulus defect measured at 4.2 nif- K in the refractory bcc metals is quite small compared with that associated with the deformation peaks along. (H.G.G.)

Separate abstracts have been prepared for 5 sections of this report. (C.H.)

A discussion of the materials requirements for seals and insulators for thermionic energy converters is presented. Based on the requirements and published physical property data, Al/sub 2/O/sub 3/ appeared to be the most desirable insulator material at temperatures below its melting point. The study includes a discussion of metal ceramic sealing technology and a bibliography of publications in the field covering 1959, 1960, and part of 1961. (auth)

Progress is reported in MSRE design, engineering analysis, component development, metallurgy, in-pile tests, chemistry, fuel processing and in thermophysical properties of salt mixtures. (J.R.D.)

This document details activities of the Irradiation Processing Department during the month of September 1962.

Since any radiography must be performed in an enclosed system, an enclosed glove box was constructed to facilitate radiographic operations. Tests show that no significant detail is lost in radiographing dense materials greater than 0.020 inch through 0.120 inch of aluminum using energies in the 1 Mev range. (auth)