Further calculations have been made to determine the desired radial fuel distribution in the HFIR and in the forthcoming HFIR critical experiment. In the process the design of the core was changed to include a 1-cm-thick annular space of water between the two nearly equally thick fuel annuli, a metal-to-water ratio in the fuel annuli of 1.0 (0.050 in. thick plates and coolant channels ) was specified, and the active length of the core was increased from 18 to 20 in. Results of the calculations indicated that the largest ratio of maximum meat thickness to minimum meat thickness occurred in the inner fuel annulus and was equal to 3.6, while the maximum fuel concentration occurred in the outer fuel annulus and was equivalent to about 0.7 g of U-235/cm^3 of meat. The total U-235 loading for this core was 8.01 kg, which results in a core lifetime of about 14 days.

The IBM-70% code GHT has been programmed for the 7090computer, to be run under control of the IBM-7090 FORTAN Monitor System. By eliminating the use of tapes for intermediate number storage, it has been possible to reduce computing time by a factor of 26 over the 704. A minimum of three tapes are required; one for input, one for output, and the System tape. One scratch tape is required for transient problems, and one punch tape is required if the steady-state temperature distribution is to be punched on cards.

Equilibrium krypton and xenon adsorption isotherms were obtained for four varieties of charcoal and for Linde Molecular Sieves Types 4A and 5A, generally at 0, 25, and 60°c. Such data are of interest in connection with design and evaluation of adsorbers for radioactive noble gas fission products. The isotherms were fitted, by linear regression analysis, to straight-line forms of the Freundlich and Langmuir equations. The Freundlich linear equation gave the better fit and the parameters of this equation are presented for each of the isotherms. Also presented are the constants for an equation representing the temperature dependence of arbitrary adsorption coefficients, the coefficients having been calculated from the Freundlich isotherm parameters. Some aspects of the applicability and accuracy of these results are discussed.

A program translates an algebraic language like Algol into the machine language of an electronic computer must perform the following functions.

Preliminary results from experiments on the dissolution of beryllium metal and sintered UO2-BeO fuel pellets are reported. In all cases the pellets were fired in hydrogen at 1650-1800°C. Uranium, from UO2-BeO pellets containing more than 60% UO2, is readily leached with boiling 6-13 M HNO3 in about 6 hr. The BeO in these pellets dissolves only slowly in nitric acid; however, in 8 M HNO3-0.2M NaF, it dissolves at about the same rate as the UO2. Sintered pellets containing less than 10% UO2 do not dissolve rapidly in common aqueous reagents. The highest rates are obtained in boiling acidic fluoride solutions; e.g., sintered BeO and BeO-8% UO2 are dissolved initially at a rate of about 1.7 mg min-1cm-2 (13 mils/hr) in boiling 5.8 M NH4HF2.

Test work was completed on development of a stripping method for the amine extraction (Amex) process which produces a concentrated uranyl nitrate solution for shipment to the refinery. This procedure offers potential cost savings by simplifying the overall mill-refinery flowsheet. The process involves treatment of the amine extract with calcium nitrate solution to convert the uranium in the solvent to a nitrate complex, stripping the uranium with water or dilute nitric acid, and recovery of nitrate from the solvent for recycle by contact with a lime slurry.

A cross section tape has been prepared containing data for elements of most interest for aqueous homogeneous reactor multigroup diffusion calculations involving combinations of 150, 200, 240, 260, and 280°C temperatures. This tape permits multigroup analysis of a reactor in which two or more regions are at different temperatures.

A multiple controller system for a single process is formulated in block diagram form. It is shown that for the general case of n controllers and one type of cross coupling, the system equations can be reduced to the standard forms of servomechanism theory. The effects of failure in individual controllers are examined, and it is demonstrated that the system shows considerable promise for both reliability and maintenance without process interruption.

THOROBRED is an IBM-704 code for the calculation of nuclear and economic characteristics of two-region homogeneous reactors operating on a U-Th fuel cycle in the steady state. The major input variables are reactor dimensions, power level, thorium concentration, chemical and purification systems cycle times, corrosion rates, fraction of Pa loss, and nuclear and economic data. The output includes equilibrium isotope concentrations, neutron balance, and a breakdown of fuel cost. Running times are normally less than one minute per case.

The equilibrium and maximum override values for xenon and samarium poisoning have been computed using the recently issued effective cross sections of C. H. Westcott. Values are given as a function of specific power, neutron temperature, and epithermal flux content.

The general solution to the basic differential equation d^4w/dy^4 +4w= -4f(y) is transformed from the primary form treated in most texts to an alternate form in which each integration constant corresponds to one edge condition at y=0. The relationships between the integration constants of the two forms are derived and the values for the transformed functions are tabulated. The particular solution is derived in general and given in unique form for various functional forms of f(y). Matrix notation is used throughout the derivations; however, a knowledge of matrix theory is not need for application of the results.

A review of helium sources, availability, consumption, and future production capacity indicates that the helium consumed in the operation of 100 helium-cooled reactors for 300-Mwe power plants over a 20-year period will run no more than a few per cent of the helium consumed in the U. S. in the next 25 years and less than 1% of the helium that is to be stored during that period under the helium conservation act of 1960 (Public Law 86-777).

Analyses are generally made at the ORR with a view toward determining the surface temperatures of fuel plates.

A listing and description is given of the experiments associated with the HFIR Critical Experiment-2. The primary experiments concern the reactivity of the bare core, reactivity worth of "gray" control plates, core-power distribution, reactivity. The secondary experiments concern the reactivity of the fuel, and the reactivity worth of a "partial" gray plate.

The absorption in lead of Bremsstrahlung X radiation from a Sr^90 TiO3 pellet in the proximity of Hastelloy "C" was measured. The tenth value layer of the more energetic components of the X-ray continuum was determined to be 1.60 inches.

A carrier containing 138.99 liters of solution, uranium concentration 202.04 g/liter with an isotopic concentration of 97.3% U-233, was prepared for shipment. The total uranium was 28,062 +/- 60 g (95% confidence level) and the U-233, 27,305 +/- 66 g (95% confidence level).

Simulated reactor fuel solutions were circulated at temperatures as high as 365°C in a small titanium pump loop. A hydroclone separator separated heavy phases formed at high temperatures. As the temperature of the solution was increased beyond the two-liquid-phase temperature (327°C), the salt concentration of the light phase decreased and the acid concentration increased. The mole ratios of uranium to sulfate, uranium to copper, and uranium to nickel in the light phase decreased in the same proportion in the temperature range of 330 to 365°C. Corrosion of titanium and Zircaloy-2 specimens was insignificant during the relatively short exposure periods.

Startup accident and pump run-down on the ORR have been simulated on the Reactor Controls Analog Facility. At full flow the 150% level scram (45 Mv) easily terminates the startup accident before the metal temperature gets above 180°F. For very low flows typical of criticality runs, temperature coefficients turn the excursion before it reaches 150% of full power, and temperatures climb to boiling, a potentially hazardous condition. (This same behavior can occur at full flow is the power is increased to the point where the level scram must be set above 50 Mw).

The experimental demonstration of fission product release from over heated reactor fuel is necessarily subject in many respects to the arbitrary conditions imposed by the experimenter. Since an almost infinite latitude exists in the choice of materials, atmospheres, gas, velocities, temperatures etc., some allowance for an extrapolation to alternate conditions is definitely in order. The conditions imposed in this study are best described as those most likely to maximize fission product release. Two of the most important variables not investigated in the present report are the influence of metal cladding and the difference in internal nuclear heating as opposed to external radiant heating. In addition a significant uncertainty exists in the understanding of diffusion through large masses such as might result from a scaled-up melt down in a reactor.

Sanderson & Porter have carried out a series of studies over the last four years which indicate that the pebble-bed reactor way be an attractive way to obtain low-cost power. At the request of the Atomic Energy Commission, two design studies have been carried out on this concept at the Oak Ridge National Laboratory. The first of these a preliminary design of a 10-Mw(t) reactor experiment, the PRRE, was initiated September 10, and a report on the study was issued November 1960. The second phase of the work, a conceptual design study of a 330-Mw (e) central station, was initiated November 1, and is the subject of this report.

Studies are being made on the recovery of thorium (and uranium) from granitic rock, since this source represents a very large potential thorium reserve for the nuclear power industry. In preliminary leaching studies on 16 granite samples (containing 8-95 ppm thorium and 1.5-16 ppm uranium), maximum recoveries of thorium and uranium ranged 30-85% and 15-65% respectively, and sulfuric acid consumption was high (30-120 lbs H2SO4 per ton of granite). A relatively high acidity was needed to obtain rapid and efficient dissolution of the soluble thorium fraction. The cost of treating granite was estimated at $3.50-5.20 per ton, variations within this range being dependent primarily on differences in acid consumption for different granites. Estimated costs per pound of thorium plus uranium recovered ranged $30-500.

The development of nuclear reactors and other sources of nuclear particles has provided the analyst with a new method which has been successfully applied to the determination of microgram and even submicrogram quantities of many elements. In this method, known as "radioactivation analysis," the element to be determined is "activated" through some type of nuclear reaction which produces a radioactive isotope of the element. Since the radioisotope produced decays with its own characteristic radiation and half-life, it is possible, through radiochemical measurements following radioactivation, to devise a method of analysis which is very specific for particular elements. Whenever necessary, the radioisotope is separated by chemical means and its radioactivity is then measured by some type of radiation counter.

The objectives of this study have been to examine the problems of the pebble-bed reactor concept and to conceive a design of a facility for investigating the feasibility of this type of reactor. The design must provide for adequate leaktightness of the contaminated-gas system and adequate maintenance of contaminated components, the most vital feasibility questions of the concept.

This manual represents the recommendations of the Instrumentation and Controls Division committee on Electrical and Electronic Symbols and Drawing that have been issued to date. Section I consists of copies of the American Standards Association Graphical Symbols for Electrical Diagrams, with certain variations or additions that have been recommended by the Committee to clarify or more positively identify the device or element symbolized. Since publication of the present ASA standards, certain elements, such as the transistor, have come into common use. Committee and represent the symbols have been selected after considerable study by the Committee and represent the symbols that seem to be standard Reference Information series, Interim E--1.