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Report giving tables of the linear contraction relative to 293 ºK, (L293ºLT)/L293, and the linear expansion coefficient, dL/L293dT, of thirty elements, forty-five alloys, twenty-two other inorganic substances and twenty plastics and elastomers in the temperature range, 0 to 300 ºK.

During a feasibility study of auxiliary shutdown systems for the Spert II and Spert III reactor facilities, salts possessing high neutron absorption cross sections were considered for use in water soluble poison injection systems, and gadolinium nitrate appeared promising. Tests were carried out in a high pressure static autoclave to determine the stability of gadolinium nitrate solution at elevated temperature. At temperatures of 300 deg F and above, the solution decomposed with formation of a white precipitate. It was concluded that gadolinium nitrate is unsuitable for use as a soluble poison in high temperature, watermoderated reactor systems. (auth)

Included are summaries of studies of: DCX-1 operation and performance; DCX-2 design and engineering; plasma physics; vacuunn arcs; ion production; acceleration and injection; plasma theory and computations; instrumentation; magnetics; cross-section measurements; vacuum systems and techniques; bell- lightning; and design and engineering. Separate abstracts were prepared for the 11 sections. (B.O.G.)

Report describing ongoing progress made at the Oak Ridge National Laboratory's Thermonuclear Division.

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GGBR), and deuterium- moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site. The maximum annual fuel yields were 1.5 mills/ kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore, in view of the uncertainties in nuclear data and efficiencies of processing methods, only these two can be listed with confidence as being able to satisfy the main criterion of the AEC longrange thorium breeder program, viz. a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for the other systems. The AHBR was judged to rank first in …

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas cooled graphite-moderated (GGBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium- moderated gascooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected with continuous processing of fuel and fertile streams. The maximum annual fuel yields were 16, 7, 4, 4, and 4.5%/yr, respectively at a fuel cycle cost of 1.5 mills/kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of 7, 1, 1, 2, and 3%/yr. At a fuel yield of 4%/yr, the costs were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore only these two can be listed with confidence as being able to satisfy the mdin criterion of the AEC long-range thorium breeder program i.e., a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for …

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

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The measurement of the ultrasonic wave velocities in a sufficient number of known directions in a single crystal makes possible the direct computation of the elastic moduli of the crystal. For specimens having linear dimensions on the order of 2 to 3 mm, phase comparison techniques were found useful. The equipment described is based on the technique developed by McSkimin. The measurement system is actually a type of acoustic interferometer adapted to small solid specimens, and it requires specialized electronic equipment which cannot be purchased off the shelf. (auth)

From Abstract: "The equipment described in this paper is based on the technique developed by McSkimin. Figure 1 shows a block diagram of the system. The oscilloscope presentation with no specimen present appears as shown in Figure 2a. At discrete frequencies, the pattern appears as shown in Figure 2b."

A loop was operated in the Brookhaven Graphite Research Reactor to determine the effect of in-pile irradiation on the corrosion of various materials by a U-- Bi solution. The loop wws fabricated of 21/4% chrome-1% Mo steel and contained, in the in-pile section, specimens of low-chrome steels, C steel, Mo, Be, Ta, and graphite. The U--Bi solution containing 869 ppm U/sup 235/ 98 ppm U/ sup 238/, 236 ppm Zr, and 346 ppm Mg was circulated at 51/4 gpm. A temperature difference of 75 deg C was maintained on the loop. The in-pile test section ran at 500 deg C and the finned cooler section at 425 deg C. The in-pile test section was exposed to a neutron flux of 4.4 x 10/sup 12/ neutrons/cm/sup 2/-sec which provided a fission density of 5.5 x 10/sup 10/ fissions/cm/sup 3/-sec. Metallographic examination indicated that the corrosion and/or erosion of the steel and graphite specimens was nil. Wetting of the specimens by the U-Bi solution was limited. Results indicate that in-pile and out-of-pile experimental results are similar and that fission fragment recoils did not contribute materially to either wetting or corrosion under the conditions imposed in this test. (auth)

Report issued by the Brookhaven National Laboratory discussing uranium-bismuth corrosion test loops. Design, construction, and operation of the test loops are presented. This report includes tables, illustrations, and photographs.

"This is the fourth Quarterly Report on Task II dealing with Methods Development for Radio-Tracing of Coal Product Hydrocarbons With Tritium. The problem of radio-chemical purity of traces is resolved by developing a new method of labeling with tritium that generates essentially no highly tagged side products or impurities."

Results are summarized for zero-power measurements carried out on the AeLPR to stady the characteristics of fuel assembly arrays, burnable poison strips, and control rods. The criteria were stable reactor operation at the design power of 3 Mwt, enough explicit and inherent reactivity for reactor operation for as long as 3 years, and enough control capacity for shutdown at any time during the core cycle. Two reference cores were used in the measurements: a final reference 40-fuel-assembly core for 3 Mwt operation and a reference 59-fuel- assembly loading for possible use at higher power levels in another plant. Aspects of design of the core and control system are discussed. (D.L.C.)

Report describing the zero-power experiments on the Argonne Low Power Reactor. It is a prototype of a natural-circulation, boiling-water-reactor nuclear power plant designed to supply the electrical power and space heat requirements of a proposed Auxiliary DEW Line radar station.

A series of experiments was proposed in which the thermal output of the EBWR is increased to 100 Mw. The increase in thermal enengy will be accomplished by inserting 32 spike elements in the initial fuel core. The fuel selected for the spike elements was a cubic solid solution in the system ZrO/sub 2/- CaO- UO/ sub 2/. The ceramic fuel was fabricated in the form of pellets by compacting a mixture of U/sub 3/O/sub 8/, CaCO/sub 3/, and ZrO/sub 2/. The pressed pellets were sintered in air at 1675 deg C plus or minus 25 deg C to form a cubic solid solution having a composition of 9.01 wt.% UO/sub 2/ (93% enriched), 9.07 wt.% CaO, and 81.92 wt.% ZrO/sub 2/. The procedures used in forming the fuel pellets are described. Following fabrication, the pellets were loosely inserted into Zircaloy-2 tubes for subsequent assembly into fuel elements. (auth)