The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)

None

Available knowledge of precision limits in S.S. accountability measurements and/or calculations by reactor and chemical processing groups is surveyed and summarized. Experienee in comparisons of reactor (production and research) calculations vs. chemical plant accountability measurements is also reported. A general tentative conclusion is that available precisions ( plus or minus 0.54 to plus or minus 0.78%) in chemical plant measurements (bulk and analytical) for fissionable material accountability is superior to the variable precision ( plus or minus 1.0 to 1l.0%) possible by calculations (nuclear and/or engineering) of power reactor systems; however, with operation and empirical experience (e.g., after two or three core loadings), it is believed that calculations for given reactors can attain acceptable precisions, e,g., less than plus or minus 1.0%. It may be proposed that fuel payments be made as follows: 90% of fuel value based on reactor calculations, an additional 5% based on dissolver analyses, and final settlement based on chemical plant material balance (product plus loss analyses). (auth)

Techniques were developed to measure the liquid density and viscosity of organic coolants over the temperature range 300 to 850 deg F. (W.L.H.)

This code finds inelastic cross-section matrix elements (transfer matrix) for hot monatomic moderator for multigroup calculations by numeric- analytic double integration of Cohen's formula. Several approximations to the actual neutron density ean be used as weight functions over the velocities of the initial groups. Modified and supplemented results are presented on binary cards and/or tape for direct input into the Argonne Transport Theory Codes or the SNG Code, or for offline output. (auth)

An experimental study was made to determine the effective shielding provided by a modern reinforced-concrete office building (AEC Headquarters building) from nuclear fall-out. Pocket ionization chambers were used for measurement of the radiation-field strength. Fall-out was simulated with distributed and point-source configurations of Co/sup 60/ and Ir/sup 192/ sources. Four typical sections were selected for study, and experiments were performed on each. These included an external wing with exposed basement walls and an external wing with a buried basement. Roof studies were made on an internal wing with a full basement and on the east end of wing A, which has a thin-roof construction. The thick-roof construction of 8 in. of concrete and 2 in. of rigid insulation covers all the building except the east end of wing A, which has 4 in. of concrete and 2 in. of insulation. (auth)

Batch processes on a scale of 250 to 300 g of uranium were developed for the production of uranyl oxinate (8quinolinate) and uranyl malonate. Both compounds are insoluble in water and were found to be suitably soluble in pitch. Uranyl oxinate was prepared by the reaction of an aqueous uranyl nitrate solution with an acetic acid solution of oxine (8-quinolirol) at about 80 deg C. Complete precipitation was accomplished by the addition of ammonium hydroxide. Yields of better than 99.5% were obtained. Uranyl malonate was prepared by the reaction of aqueous solutions of sodium malonate and uranyl nitrate at about 80 deg C in 97 to 98% yield. Uranyl 2-ethylhexanoate was prepared by a transesterification reaction from uranyl acetate and 2-ethylhexanoic acid. Yields of 90% were obtained but the process was quite laborious ard time consuming. A metathesis method of preparation was not successful. (auth)

Ultrasonic and eddy current methods were used to inspect EBR-I Mark III fuel elements and componentsUltrasonic techniques were used to inspect for homogeneity of the casting, bonding of the core to the clad on the extruded rod, bonding of the Zircaloy spacer disk to the uranium, and cracks in the Zircaloy rod used for end caps. Eddy current techniques were used to measure the cladding thickness on the extruded rods and to inspect the zirconium wire used for spacers on the completed fuel element. (auth)

None

None

The purpose of this production test is to provide a standard method of obtaining scram transient reactivity information at the eight reactors, under conditions conducive to valid data. These conditions include the bypassing of the Panellit system at a low power level for a short, controlled period of time during May 1959.

Results of four experimentul runs in the Fluorox fluidized bed reactor system are reported. The engineering feasibility of UF/sub 6/ production from UF/ sub 4/ by use of dry air of O/sub 2/, 2UF/sub 4/ + O/sub 2/ = UF/sub 6/+ UO/sub 2/ F/sub 2/, in an Inconel fluidized bed reactor at 800 to 850 deg C was demonstrated in two experimental tests in which greater than 90% of the theoretical amount of UF/sub 6/ was collected or measured. Two runs made with crude UF/sub 4/ (produced from unpurified mill concentrate) as the feed material, showed thnt UF/sub 6/ could be produced at 700 to 725 deg C but corrosion on Inconel was prohibitive. (auth)

None

An attempt is made to present available information pentinent to reactor containment. This is done directly, by summary and reference, or by reference alone. To provide a reference framework, the first review document must necessarily be handled differently from supplemental periodic reviews. The plan is to: (3) provide a detailed account of the problem and suggestions for work needed to yield adequate solutions; (2) present the accumulated knowledge and accomplishments; (3) give an account of experience in applying the containment concept; and (4) append extensive bibliographical material. An attempt is made in each case to indicate the significance of the information and its relation to the problems outlined. (A.C.)

The Purpose of the work described in this report has been to determine experimentally the rate of production of tritiun in fringe lithium-aluminum alloy loadings with the degree of precision necessary for economic analyses of such a method of isotope production. These results are provided for use in such an analysis.

please delete the no. 17076<><DSN>13:017077<ABS>The P/sub 1/ multigroup code was written for the IBM-704 in order to determine the accuracy of the few- group diffusion scheme with various imposed conditions and also to provide an alternate computational method when this scheme fails to be sufficiently accurate. The code solves for the spatially dependent multigroup flux, taking into account such nuclear phenomena is slowing down of neutrons resulting from elastic and inelastic scattering, the removal of neutrons resulting from epithermal capture and fission resonances, and the regeneration of fist neutrons resulting from fissioning which may occur in any of as many as 80 fast multigroups or in the one thermal group. The code will accept as input a physical description of the reactor (that is: slab, cylindrical, or spherical geometry, number of points and regions, composition description group dependent boundary condition, transverse buckling, and mesh sizes) and a prepared library of nuclear properties of all the isotopes in each composition. The code will produce as output multigroup fluxes, currents, and isotopic slowing-down densities, in addition to pointwise and regionwise few-group macroscopic cross sections. (auth)

Performance characteristics of two activated charcoal columns at room temperature in separating fission-product xenon from an air stream were investigated by installing each column in the exhaust from an enclosure in which irradiated slugs were dissolved. Breakthrough curves are presented and the variation in xenon concentration within the columns is examined. Theoretical treatments of adsorption columns in the literature are found to agree well with the experimental data. Performance of the colunms is evaluated in terms of concentration factor'' and number of effective theoretical plates. (auth)

Detailed descriptions are given concerning the reactor design, operation, location, and the using agency. Potential hazards are evaluated along with safeguards. A supplement is included which further describes the reactor core. (J.R.D.)

A general numerical procedure, called the discrete S/sub n/ method, for solving the neutron transport equation is described. The main topics relate to the derivation of suitable difference equations, and to the problem of solving these, while maintaining generality, accuracy, and reasonable computing speed. A few comparisons with other methods are made. (auth)

Considerable production gains might be attained if each reactor discharged its entire flattened region during one scheduled outage instead of utilizing several outages for this purpose. Several of the older reactors are now discharging a high percentage of their flattened zones in a single outage and could be put into this type of operation with relatively little difficulty. Production gains may be possible through better flattening efficiency, a more favorable rupture rate effect, fewer non-equilibrium losses, higher conversion ratio, and more efficient usage of outage work. Since this document is written Primarily from the Operational Physics standpoint, some gains and pitfalls which must be evaluated by other affected groups will only be mentioned here as possibilities. The purpose of this document is simply to point out the potential gains in flattening efficiency from this method. Potential gains from improved fuel performance have been described in another document.

The proposed operating conditions for fuel elements to be charged into the NPR require the fuel to be of an extended surface geometry and maintain adequate strength and corrosion resistance in 300 C water. A contract was let to Nuclear Metals Inc. to produce by co-extrusion lengths of fuel rod containing both natural and 1.6% enriched uranium of irradiation quality for fabrication into fuel elements. The fuel rods used in the irradiation test represent the first enriched uranium rods coextruded in 0.030 inches of Zircaloy-2 to be irradiated and examined at Hanford. The rods used for this test were fabricated into four, 4 rod cluster fuel elements thus allowing adequate space between individual rods for expansion in the case of a fuel rod failure. This rod was of particular interest since it contained an irregular uranium-Zircaloy-2 interface. The purpose of the irradiation was to determine the dimensional stability of coextruded fuel rods and to determine whether the irregularity in the bond interface had any effect upon the irradiation performance of the fuel. Fuel elements were irradiated in 200 C water in the KER Loop 2 facility to an exposure of 0.28 a/o burnup (2,200 MWD/T). Post irradiation examination showed that …

The results of small exploding-wire studies were found to be capable of direct extrapolation to larger wires (an increase in cross-sectional area of 1500 to 1500 from the small wires). Copper wires up to 40 mils in diameter and iron wires to 62 mils in diameter were studied for use as fuses. in lengths up to 18 in. A dependency between pause time (the time between system current cut-off and current restrike) and wire length is described for several sizes of copper wires exploded with 16.5- and 49.5-kilojoule sources. The role of wire confinement is discussed in connection with establishment of the pause. (auth)

The construction and operation of the plasma injector, Toy Top, used ia the magnetic high compression experements in progess at the Lawrence Radiation Jab. at Livemore are described The essential part of the injector consists of a stack of deuterated titanium washers 3/4 in. O.D. and/2 in. I.D. Details of the construction are sbown (W.D.M.)

The demand plot has a 5-set, modified pump decay curve; it shows that 20,000 gpm emergency flow would be required within 80 seconds of complete pump power failure. Bases for the demand curve are constant bulk inlet temperature of 2 C, constant bulk outlet temperature of 95 C, K-3 I&E fuel elements, and initial reactor flow of 188,000 gpm.