Using a special reactive form of high-purity alumina, claddings were applied to UO/sub 2/ particles by a tumbling technique. The clad pellets were isostatically pressed at 100,000 psi and then sintered at 2800 deg F in hydrogen. crack-free spheroidal pellets ranging from 1000 to 2000 mu in diameter were produced. The dense Al/sub 2/O/sub 3/ envelopes surrounding the UO/sub 2/ particles were estimated to be 300 to 500 mu thick. The Al/sub 2/O/sub 3/ claddings protected the UO/sub 2/ from oxidation when the pellets were heated in air for 100 hr at 1200 or 1800 deg . There was no measurable release of fission products from irradiated clad particles heat treated at 1700 deg F in vacuum for 7 days after exposure to 6.0 x 10/sup 12/ nv for 1 hr at room temperature. Claddings of other oxides, such as Beo or MgO, probably could be applied by the same techniques used in applying the Al/sub 2/O/sub 3/ claddings. (auth)

Some calculations have been made to estimate the characteristics of a neutron-burst type fast reactor similar to Godiva but made up of relatively small component parts--the so-called "layered assembly." One spherical and three cylindrical assemblies have been considered. Critical masses, assuming 5% voids, range from 58 to 65 kg of 93.4% enriched U/sup 235/. For a reactivity addition of 0.33 dollars above prompt criticals bursts between 2 x 10/sup 17/ and 6.7 x 10/ sup 17/ fissions were computed with accompanying temperature rises varying from 514 to 1600 deg C. The burst width at half-maximum was about 12 microseconds. To obtain an idea of the possibilities of stress reduction which might be achieved by layerings an assembly made of small rings was considered. While the critical masses obtained here are believed to be fairly accurates the predictions concerning mechanical energy generated, total fissions, and burst width may be subject to sizeable error due to the many simplitications required to allow hand computations. Neverthelesss considerable improvement in safety and burst-size is indicated by the use of a "layered assembly" instead of an assembly composed of relatively thick parts. (auth)

This annual report of Brookhaven National Laboratory describes its program and activities for the fiscal year 1958. The progress and trends of the research program are presented along with a description of the operational, service, and administrative activities of the Laboratory. The scientific and technical details of the many research and development activities are covered more fully in scientific and technical periodicals and in the quarterly scientific progress reports and other scientiflc reports of the Laboratory. A list of all publications for July 1, 1957 to June 30, 1958, is given. Status and progress are given in fields of physics, accelerator development, instrumentation, applied mathematics, chemistry, nuclear engineering, biology, and medical research. (For preceding period see BNL-462.) (W.D.M.)

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A description of the Argonne 60-in. cyclotron along with its performance, operstional characteristics, and housimig with its associated facilities is presented. (J.E.D.)

Dogs, pigs, rabbits, guinea pigs, and mice were exposed to nuclear detonatiors in two open underground pantitioned shelters. The shelters were of similar constructions and each was exposed to separate detonations. Each inner chamber filled through its own orifice; thus four separate pressure enviromments were obtained. An aerodynamic mound was placed over the escape hatch of each structure to determine its effect on the pressurecurve shape inside the chamber. In one test a sieve plate bolted across the top of the mound was evaluated. Wind protective baffles of solid plate and of heavy wire screen were installed in the shelters to compare primary and tertiary blast effects on dogs. The shelters also contained static and dynamic pressure gages, radiation detectors, telemetering devices, and, in one test, air-temperature measuring instruments, dustcollecting trays, and eight pigs for the biological assessment of thermal effects. One dog was severely injured from tertiary blast effects associated with a maximal dynamic pressure (Q) of 10.5 psi, and one was undamaged with a maximal Q of 2 psi. Primary blast effects resulting from peak overpressures of 30.3, 25.5, 9.5. and 4.1 psi were minimal. The mortality was 19 per cent of the mice exposed to a peak …

Dogs, pigs, rabbits, guinea pigs, and mice were exposed to nuclear detonations in two open underground partitioned shelters. The shelters were of similar construction, and each was exposed to separate detonations. Each inner chamber filled through its own orifice; thus four separate pressure environments were obtained. An aerodynamic mound was placed over the escape hatch of each structure to determine its effect on the pressure-curve shape inside the chamber. In one test a sieve plate bolted across the top of the mound was evaluated. Wind protective baffles of solid plate and of heavy wire screen were installed in the shelters to compare primary and tertiary blast effects on dogs. The shelters also contained static and dynamic pressure gages, radiation detectors, telemetering devices, and, in one test, air-temperature measuring instruments, dust-collecting trays, and eight pigs for the biological assessment of thermal effects. One dog was severely injured from tertiary blast effects associated with a maximal dynamic pressure (Q) of 10.5 psi, and one was undamaged with a maximal Q of 2 psi. Primary blast effects resulting from peak overpressures of 30.3, 25.5, 9.5, and 4.1 psi were minimal. The mortality was 19% of the mice exposed to a peak pressure of …

The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)

This report discusses the events from a chemical standpoint following a total loss of coolant disaster will not be altered in the melting reactor by the introduction of N metal. The interdiffusion of uranium and aluminum will be the dominating reaction, causing the blockage and tying up of the lithium in UAl{sub 3} which does not melt until after the uranium does. Pressure from the swelling UAl{sub 3} will extrude uranium-aluminum and lithium into graphite weep holes and block interfaces. The migration of lithium by vaporization will not became appreciable until well over 2000{degrees}C, well beyond the time when uranium and UAl{sub 3} have melted. The eventual result will be a diffuse distribution of uranium, lithium, and aluminium in the lattice. The E-N pile has a larger excess over required control capacity than the uranium provided the large reactivity poison tied up in the lithium is not lost. Compared to the natural uranium pile, the gain of reactivity on loss of coolant is less and the net temperature coefficient in the dry pile remains negative to higher exposures. Furthermore, permanent pile poisoning during meltdown is accomplished via two mechanisms both lithium and uranium redistribution in the lattice produce large negative …

This report for January 1959, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.

It was demonstrated that neptunium(IV) can be readily absorbed onto anion exchange resins from 6 M HNOsub 3/ containing ferrous sulfamate and hydrazine or semicarbazide, separated from plutonium, uranium, and common metallic impurities by washing the resin at 25 deg C with 6 M HNO/sub 3/ containing ferrous sulfamate and hydrazine or semicarbazide, separated from fission products and thorium by washing the resin at 60 deg C with S M HNO/sub 3/- 0.01 M HF containing hydrazine or semicarbazide, and eluted at concentrations greater than 40 g Np/l with 0.35 M HNO/sub 3/ at 25 deg C. Decontamination factors of greater than 10,000 from uranium, plutonium, and common metallic contammants, greater than 25,000 for fission products normally expected in the feed (mainly Zr-Nb with some Ru-Rh), and greater than 1000 for thoriuin are obtainable under proper operating conditions. Because of the low processing rates, the necessity for carrying out the absorption cycle at 25 deg C and the absence of radiation daraage problems, Dowex 1, X-4 (50-100 mesh) or Dowex 21K (50-100 mesh) resins are considered the best choices for this application. Gassing occurs with the use of ferrous sulfamate - semicarbazide reductant but is not a serious problem and …

A study to define the effectiveness limits of sulfamic acid and to discover other better stabilizers for Pu(III) is described. Ascorbic and isoascorbic acids, used in conjunction with sulfamic acid reduced Pu(IV) to stable Pu(III) in nitric acid. Aminoguanidine sulfate also retarded the oxidation of Pu(III) but did not reduce Pu(IV). (J.R.D.)

(((Abstract unscannable)))<><DSN>13:014503<ABS>Thc Nb-O equilibrium system was determined by metallographic examination of arc-cast alloys made ot electron-gun-refined Nb metal and special purity Nb/sub 2/O/sub 5/. Two intermediate oxides. NbO and NbO/sub 2/, melt without decomposition at 1945 C and 1915 C, respectively. Eutectic reactions exist between Nb and NbO at 1915 C and between NbO and NbO/sub 2/ at 1810 C . Experimental evidence supports a peritectic reaction between NbO/sub 2/ and Nb/sub 2/O/sub 5/ at 1510 C. The maxinium solid solubility of 0 in Nb metal is 0.72 wt.%. (auth)

It was found from graphite irradiations that the changes can be described by a growth component and a contraction component, and that the contraction is a linear function of exposure, while the growth is a function of both exposure and temperature which saturated at high exposures, say 3000 MWD/a ton. A graph is included. Contraction rate is determined for KW, KE, and H piles.

BS>Critical concentration calculations were made for several D/sub 2/O-H/ sub 2/O moderated HRT-type reactors with 30- and 28-in. core diameters and pressure vessel diameters of 60 and 54 in. A core temperature of 300 C was assumed for all cases while the blanket temperatures assumed the values 250, 280, and 300 C. The assumed moderator compositions were 80, 90, and 100% D/sub 2/O. (auth)

An estimate of the critical current for the case of breakup by an arc has been obtained. It is shown that the same technique allows a quick estimate of the critical current in the case of no arc, using neutral and plasma breakup instead. (auth)

Based on the results of detailed fabrication studies, an evaluation of the effects of varying the type and size of UO/sub 2/ particles, the type and size of stainless steel matrix powders, blending procedures, compacting pressures, sintering times, temperatures, and atmospheres, roll-clading temperatures and reduction rates, total cold reduction, and heat-treating times and temperatures was made for UO/sub 2/stainless steel dispersion fuel elements. Transverse tensile tests, creep-rupture tests, metallographic examination, radiography, density measurements, and x-raydiffraction studies were used to evaluate the structure and properties of the fuel elements. From these studies a reference fabrication procedure for GCRE fuel elements was established. The fuel element core contains minus 100 plus 200-mesh hydrothermal UO/sub 2/ dispersed in an 18-14-2.5 alloy matrix prepared from minus 325-mesh elemental iron, chromium, nickel, and molybdenum powders. Commercial Type 318 stainless steel is used for cladding. Core compacts are sintered in steps to 2300 deg F after cold compacting at 15 tsi. Evacuated picture-frame packs are hot rolled from a hydrogen muffle at 2200 deg F with a 40% reduction in thickness on the first pass and a 20% reduction in thickness on remaining passes. After annealing at 2300 deg F, the fuel elements are given a …

ABS>Published information is not available for accurate prediction of the electrical insulating characteristics of helium at high pressures and temperatures. In general the breakdown voltage increases as the gas pressure is increased and decreases as the gas temperature is increased. The relatively low breackdown voltage of helium accents the importance of additional investigation in this field. (auth)

The two predominate methods of dispersing uranium in graphite are reviewed and evaluated. This study indicated that the most feasible method of dispersing uranium in graphite would be to fabricate a mixture of graphite and U/ sub 3/O/sub 8/ bonded with a thermosetting resin. A commercial type graphite was developed through independent research, and this fabrication procedure was adapted for the manufacture of the TREAT fuel matrix. (auth)

Following-the K-pile start-ups in early 1955, a program of planned power raises was begun. The operating level had reached 1700--2000 MW by late 1955, and a severe operational control problem became apparent; the power distribution in the reactor was difficult to control and appeared inherently unstable. A study of available data led to the initiation of a production test so that a more detailed study of the phenomena could be made. This report describes the measures taken which led to an improvement in the operating characteristics of the K-piles; the current status and future outlook are also discussed in a general way.

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)

This document details activities of the Fuels Preparation Department during the month of January 1959. (FI)