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This report provides for each process tube the coolant flow rat (gpm) and temperature drop ({degrees}C) for the process tube containing the fuel rod bundles and for the four nearest neighboring process tubes. Reactor shutdown and startup times are also given. Peak thermal neutron flux in the adjacent fuel can be calculated.

Some of the controllable variables have been investigated in casting of lead alloys having 0.06 to 0.09 wt.% Ca. The alloy has been found to respond to solution heat-treatment, room-temperature precipitation hardening, and accelerated hardening at 100 ts C. Tensile strengths to 7000 psi have been obtained. (auth)

Temperatures were calculated relative to the storage of radioactive solid waste as a function of time and radial distance for radioactive solid cylinders in infinite solid media of "average soil," "average rock," and salt. A resistance at the cylinder--infinite medium boundary was included in the form of an air space. For the range of parameters used and withia the practical limits of accuracy, the maximum temperature rise increased linearly with the heat generation rate. The fission product spectrum was not significant in the determination of the maximum temperature rise. Under the pessimistic storage conditions assumed, the storage of cylinders of a practical size appears feasible without excessive temperature rise. A maximum temperature rise of 1000 deg F would be produced with an initial heat generation rate of 1300 to 1600 Btu/hr-ft/ sup 3/ for cylinders with a 5-in. radius, with 350 to 450 Btu/hr-ft/sup 3/ for a 10-in. radius, and with 175 to 210 Btu/hr-ft/sup 3/ for a 15-in. radius, assuming a thermal conductivity of the radioactive cylinder of 0. 1 Btu/hr-ft- deg F. (auth)

The computer program FATT for calculating the transient temperature distribution in the EGCR core is described. The program, in FORTRAN for the IBM 7090, provides an arbitrary choice of axial increments for the reactor core. The heat transfer coefficients are made temperature-dependent. Provisions are available for plotiing the results from the program on the Calcomp plotter. The equations and some typical results are included, and the finlte difference method of solution is described. (auth)

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The basic heat transfer and flow equations applicable to studies of flow oscillations in NERVA engines are first linearized, then expressed as difference-differential approximations, and finally written in a form suitable for analog computer solution. Block diagrams are presented which illustrate on analog procedure for solving the linearized, difference-differential equations.

The design, performance, and cost information on the nuclear portion of the Mixed Spectrum Superheater power plant are emphasized. The research and development programs required to ensure plant feasibility are also presented. The nuclear steam supply system, reactor auxiliary systems, radiation control systems, control and instrumentation, special test instrumentation, plant operation and maintenance, steam cycle, turbine plant, general service systems, preliminary safeguards considerations, expansion of plant power output to 150 Mw(e), and MSSR critical experiment are described. (M.C.G.)

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A system of six codes is described that permits the solution of a broad class of problems encountered in the study of neutron thermalization, average cross sections, and spectral models. These programs are all compatible with use on MONITOR for the IBM 7090. The Fortran listings are given. (auth)

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A transfer function for d-c operational amplifiers with generalized input and feed-back networks is derived by application of network theory. This transfer function, more general than others in common use, applies to three and four-terminal networks. It contains, as a special case, the well-known transfer function for d-c operational amplifiers with two-terminal networks. Examples are given. (auth)

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A test to obtain data for use in determining the reactivlty wohh of Na in the Hallam reactor core is described. The test is designed to obtain information on the dry temperature coefficient of reactivity and to train operators. An evaluation of results is included. (J.R.D.)

Station 1A was the emplacement hole for the nuclear device used in the Advanced Research Project Agency's and U.S. Atomic Energy Commission's 5-kT nuclear Salmon Event of October 22, 1964. The hole penetrated the Pascagoula and Hattiesburg Formations, undifferentiated, and the Catahoula Sandstone, all of Miocene age, before entering the caprock at a depth of 882 feet. In continued through the calcite, gypsum, and anhydrite units of the caprock, which extend to a depth of 1,469.5 feet, and terminated in the salt stock at a depth of 2,802 feet. Five sand aquifer units lie above the caprock, whereas one sand aquifier lies within the calcite unit of the caprock. Drilling activities did not cause noticeable water-level fluctuations in the sand aquifers; however, during drilling in the calcite caprock, drilling fluid losses twice caused significant but temporary rises in water level in three observation wells that monitor the calcite limestone aquifer. The microscopic examination of drill cuttings collected at 10-foot intervals to a depth of 1,220 feet in Station 1A is presented. 2 figures, 1 table.

This report gives the exposure data for the C Reactor at the Hanford Reservation February 1965.

During the 18-year history of Hanford operations, a high degree of safety has been provided at reasonable cost by the storage of high-activity liquid wastes in underground tanks, by the percolation of low-activity liquid wastes (principally water) through the soil to the ground water, and by the adsorption or filtration of radioactive materials from gaseous effluents prior to their discharge to the atmosphere. Studies of the Hanford area indicate that current practices could be continued almost indefinitely without jeopardizing environmental safety, provided the stored liquid wastes are periodically transferred to new tanks before the existing ones fail. In 1960, a Fission Product Recovery Program was prepared outlining a plan to meet the Atomic Energy Commission`s needs for isolated fission products. Emphasis was placed on developing processes for isolating selected fission products and on providing interim production capability for recovering the fission products of immediate interest. In 1961, development efforts were channeled almost exclusively toward meeting the immediate needs of the Commission for separated strontium-90 and cesium-137. During this time, it became increasingly evident that substantial economies could be realized, both in immediate development and subsequent operating efforts, if the separate fission product recovery and waste management programs were merged. The …

It is recommended that the ``square pile`` concept be adopted for all disaster total control calculations, and that the basic reactor constants listed in HW-62884, except for Ball 3X local strength at the DR Reactor, be used in applying this method. Curves are included for each reactor type, indicating allowable enrichment based on appropriate local control strengths. (The reactors whose operating methods are affected by disaster total control requirements are B, D, F, and DR Reactors; the remaining piles have sufficient geometrical coverage). An example of the analytical method is included.