Cratering is a subject that has been studied by many investigators for many years for many purposes. These purposes range from experimental studies of physical properties to large scale excavations using explosive charges of kiloton size. In the past ten years considerable effort has been devoted to cratering experiments for the purposes of determining the effects of cratering by nuclear explosions, with recent accent on Plowshare applications. From the large amount of data available for craters in alluvian has been possible to establish very reliable relationships between charge size, depth of bursty crater radii, and crater depths. In addition it has been possible to construct a preliminary theory of the mechanics of explosive crater formation. The available experimental data for nuclear and high explosive craters are reviewed, with particular emphasis on the data for desert alluvium, and the pertinent relationships are derived. A theory of the important cratering mechanisms, which has been evolved on the basis of these data and data from other sources, is outlined. (auth)

As the first part of a study to evaluate the economics of the various steps leading to and including the permanent disposal of high-activity liquid and solid radioactive waste, costs of interim liquid storage of acid and alkaline Purex and Thorex wastes were estimated for storage times of 0.5 to 30 years. A 6- ton/day plant was assumed, processing 1500 tons/year of uranium converter fuel at a burnup of 10,000 Mwd/ton and 270 tons/year of thorium converter fuel at a burnup of 20,000 Mwd/ton. Tanks of Savannah River design were assumed, with stainless steel construction for acid wastes and mild steel construction for neutralized wastes. The operating cycle of each tank was assumed to consist of equal filling and emptying periods plus a full (or dead) period. With interim storage time defined as filling time plus full time, tank costs were minimum when full time was 40 to 70% of the interim storage time, using present worth considerations. For waste storage times of 0.5 to 30 years, costs ranged from 2.2 x 10/sup -3/ to 9.5 x 10/sup -3/ mill/kwh/sub e/ for acid wastes and from 1.7 x 10/sup -3/ to 5.1 x 10/sup -3/ mill/kwh/sub e/ for neutralized wastes. …

Preliminary design concepts are presented for a wire scanner for experimentally evaluating spatial variations of neutron flux in the SM-l reactor core. Results of a literature search and determination of optimum criteria for flux mapping the core in minimum time dictated requirements for design concepts and specifications. The utility of both manually instrumented and automatically instrumented wire scanners was analyzed with respect to rapidity of measurement, selectivity of detector location, cost, value of data, plant downtime, and additional factors. (auth)

The Idaho Chemical Processing Plant for several years has been studying various pulsing systems, used in extraction columns for processing irradiated nucIear fuel elements, in an effort to eliminate the mechanical difficuittes in present pulsing systems. An air puising system has been demonstrated, and a discussion is given of experimental work, pulser design, and plant operational experience with this system. It is concluded, from work completed thus far that air puising of extraction columns is a considerable improvement over other systems. (auth)

Binary mixtures of BeO with the oxides Al/sub 2/O/sub 3/, CaO, Li/sub 2/ O, MgO, Sc/sub 2/O/sub 3/, TiO/sub 2/, Y/sub 2/O/sub 3/, and ZrO/sub 2/ were fired to temperatures in excess of 2050 deg C in an attempt to produce a stabilized cubic crystalline modification of BeO. No evidence was observed in microscopic and x-ray diffraction analyses of the cooled specimens that a cubic form of BeO had formed in the experiment. (auth)

Presented at the Eleventh Professional Divisions Conference of the San Francisco Section of the American Society of Mechanical Engineers, Noveraber l6, l96l. The MGCR program and its objectives are discussed. The basic MGCR plant is described. The design of the Experimental Beryllium Oxide Reactor is also described. (M.C.G.)

Available pressure data on the B and D downcomers and approach piping is tabulated herein. All readings were taken with an air bubbler system; readings given in psi were taken on a Bourdon-type pressure-vacuum gauge, readings given in inches of water were read on an oil-filled manometer. In each case, fluctuations occurred and the mean value was estimated by eye. Data are presented as a function of flow rate and temperature.

The original test authorized the discharge of three columns of irradiated fuel elements from KE Reactor as a first step in an Operational Recharging program. Purpose of this supplement is to authorize discharge of any number of columns of irradiated fuel elements after shutdown, and to continue in force Supplement A, with changed timing of operational discharge.

Irradiation of the initial test in this program involving ten tubes of alternately charged nickel-plated C-64 alloy clad test elements and X-80001 alloy control elements has been successfully completed. The test indicated that the nickel-plate spalling problem has been resolved, as no significant spalling or flaking was observed during the post-irradiation examination. The second test in this program will be to verify that the nickel-plate integrity problem has been solved by irradiating a pilot loading (up to 100 charges) of fuel elements which have been nickel-plated on a production basis.

The Overbore Test Facilities of the C Reactor are to be expanded to permit pilot testing for the complete conversion of the reactor to enlarged fuel channels. This report sets forth the technical bases and appropriate criteria to accomplish this work.

A calculation of tritium beta energy absorption in spherical bulbs filled with mixtures of tritium and other gases is ' carried out under the empirically substantiated assumption that the energy absorption per spherical shell surrounding a point tritium source is representable by a pure exponential function. Primary and successive backscattering at the wall of the vessel is taken into account by means of a very approximate calculation. The computed energy absorption function for hydrogen gas in a silvered bulb agrees within plus or minus 3% over a wide range of gas pressure with the one available set of spherical ionization chamber data. (auth)

Pu nitrate production was above revised forecast. 141 kCi Sr-90 was isolated in Purex and stored for decay; to date, 1190 kCi has been recovered from Purex wastes. Casks of Cs-137 and Sr-90 were turned over to AEC for ORNL. Production of UO{sub 3}, Pu metal buttons, and weapon component fabrication exceeded forecasts. Purex processing continued until April 14. The Purex neptunium campaign had low yield because of loss to waste system. The cartridge in the Purex co-decontamination column scrub section was replaced with a new design. After the Redox organic distillation pot was replaced, {gamma} activity in Redox recovered hexone increased 300 {times}; product quality returned to normal after the faulty steam traps were replaced. Fabricated parts underwent dimensional changes after machining and gaging; part dimensions have been altered to compensate. Design for the proposed caisson-lateral leak detection system was revised to contain 3 horizontal laterals beneath each tank and 7 vertical wells around each tank. A ratio exists between the Cs-137 in the coating waste solution and the total U in liquid and precipitate phases. (DLC)

A series of pilot plant runs was conducted to define new cartridges for increasing the capacity in the Purex 1 Bx, 2A, lC and 2E columns and eliminate plastic cartridge failures in the HA column scrub section, the HS column and the 2A column. The most favorable designs are presented and data from the various runs are included. (J.R.D.)