Results are reported from a series of experimental highexplosive shots under inversion conditions at the Nevada Test Site which were made in an attempt to refine blast prediction techniques. Applications of the data in determinations of the amount of energy which remains in the blast wave as it reaches acoustic level and in determinations of the magnitude of the reflection factor when the blast wave strikes the ground are discussed. Data on shock wave propagation are presented graphically. It is concluded that the blast phenomenology of high-altitude shots can be predicted by using modified Sach's scaling. With some extrapolation to the height-of-burst versus blast-yield curve, it should be possible to make order-of-magnitude predictions of blast effects from high-altitude shots up to heights of burst of 1,000,000 ft. (C.H.)

In view of the more-or-less eminent conversion of at least some of the older Hanford reactors to minimum process tubes it has been requested the feasibility and economics of increasing the process channel size by overboring be studied. This report is concerned with the feasibility from an operational physics standpoint of either raising reactor power levels with present aluminum process tubes and redesigned fuel elements raising reactor power levels with zirconium replacement tubes of current outside diameter, or raising power levels with either zirconium or aluminum tubes of 200 mils greater outside diameter. This report is of a survey nature only, and the data contained herein should be considered in that light.

Abstract: Testing of the Natural Circulation Steam Generator has been conducted with the Sodium Reactor Experiment. A necessary modification of the original feed-water control system was utilized.

A 1.6% enriched U rod was irradiated in 105 DR to 450 MWD/T, using a NaK-filled Al capsule. The capsule was discharged in November 1958 and examination begun. Results showed that the NaK in the capsule was trapped in the expansion chambers, leaving part of the fuel rod uncovered, causing gross heating on one side and failure of the rod, preventing the desired data from being obtained.

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The operational hazards associated with the use of an isotope-fueled auxiliary power unit for a satellite mission are described. The effects of missile about on the generator are discussed. The generator design is described, and the properties of the various fuel forms are investigated. The characteristics of the fuel capsules and the provisions for biological shielding are also described. Integration of the generator into a typical missile system is discussed. Hazards and procedures of transporting and handling the fuel cores from fabrication to launchlng are considered. Aborted missions are defined, and the forces acting on the generator during abort are described. (W.D.M.)

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lattice parameters and space-dependent neutron spectra. The method follows neutron histories through many generations, the total number depending on the statistical uncertainty desired. The major differences in the proposed Monte Carlo scheme and previous Monte Carlo calculations are enumerated. The use of neutron weights is introduced into the routine for the calculation of the fast fission factor. The calculation of the resonance escape probability employs the use of doubling surfaces in the moderator. In computing the thermal utilization, the thermal neutron spectrum is not assumed to be in the usual Wigner-Wilkins form but is calculated for every region. The effects of spectral hardening and epithermal fission are included. The neutron flux is calculated according to the definition that the neutron flux is proportional to the sum of the path lenthhs per cubic centimeter. The proposed method is applicable to computing thermal neutron spectra in both homogeneous and heterogeneous lattices. (C.J.G.)

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A series of hydrodynamic tests was conducted on a 1: 12 scale mcdel of the Yankee reactor vessel and core with four, three, and two loop operation. There was no depression of flow in the center of the core due tc the control rod shrouds or shroud support plate. With four loops operating, flow distribution tests at the core inlet indicated that for all operating conditions the highest velocities occurred in the center region of the core and were a maximum of 18% higher than the average velocities. The lowest velocities were in the region near the core baffle and were a maximum of 18% lower than the average velocities. The effective flow starvation at the points of high flux were less than the 7% design value. Mixing studies indicated that no inlet position of the core received more than 55% of its total now from one loop. The flow distribution did not change appreciably for four, three, or two loop operation. The measured pressure drop across the vessel correlated within 6% of the predicted value. (auth)

In a continuing program, fabrication characteristics, physical and mechanical properties, and corrosion behavior in air, CO/sub 2/, NaK, water, and steam were studied for . binary niobium fuel alloys containing 10, 20, 30, 40, 50, and 60 wt.% uranium To evaluate the effects of two major impurities of niobium, oxygen, and zirconium, three niobium base stocks, differing according to the level of these impurities, were used for each alloy. The impurity combinations employed were 600 ppm oxygen and 0.74 wt.% zirconium, 700 ppm oxygen, and 0.17 wt.% zirconium, and 300 ppm oxygen and 0.02 wt.% zirconium, Representative specimens of these alloys retained their hardness up to 900 deg C The 10 and 20 wt.% uraniuin alloys were successfully rorged at 2500 deg F and rolled at 1800 deg F to sheet. Fabrication characteristics of the remaining alloys are under investigation. The 0.2% offset yield strength of the 10 wt.% uranium alloy was 57,200 psi at room temperature and 36,900 psi at 1600 deg F. For the 20 wt.% uranium alloy it was 93,200 psi at room temperature and 71.000 psi at 1600 deg F. The corrosion life of all of the alloys in air at 572 deg F and in …

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; D = > 6 ; : 6 < 7 8 9 7 9sion theory calculation of the power-density distribution in the EGCR was made in order to reduce the uncertainties in the results of previous calculations. A comparison was made of calculated neutron flux distributions in seven-rod fuel clusters and the preliminary results obtained in the Physical Constants Test Reactor at Harford. Neutron flux ratios for the EGCR lattice cell were calculated for fuel enrichments of 2.0 and 2.6%. Studies were made of the power densities attainable in gas-cooled reactors operated with ceramic material as both fuel and moderator. Extensive studies were conducted to determine how the multiplication factor of a gas-cooled reactor varies with the number of rods in the fuel element cluster, cladding thickness, cladding material, inter-rod spacing, lattice pitch, solid and cored rods, fuel enrichment, and fuel-tomoderator ratio. Reactor Design Studiesi A theoretical study is being conducted of the deflections and stresses in fuel element cladding based on arbitrary temperature distributions. Tests were run to determine the load-carrying ability of the threaded-pin-type graphite column joints proposed for the EGCR. A test program was initiated for studying the behavior of metal-clad graphite bodies. An analytical model for …

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