Studies were made to ascertain what useful satellite and/ or space missions may be accomplished with the SNAP 2, 8, and 10 power systems within the 1960 to 1970 time period, to delineate useful satellite and/or space missions which would most profitably employ nuclear power sources, to ascertain what restrictions are imposed on various payloads as a result of having a nuclear power system on board, and to investigate the general vehicle integration and installation requirements of the SNAP 2 and 10 power units. Space flight programs, characteristics and limitations of SNAP, applications of nuclear power supplies. design integration, and system integration are discussed. (M.C.G.)

From introduction: The study is based entirely on observations which have been made at a number of different installations at which boiling songs and vibrations were noted during heat transfer experiments.

Corrosion screening tests were carried out on potential materials of construction for use in a fused salt-liquid metal extraction process plant. The corrodents of interest were NaCl--KCl-- MgCl/sub 2/ eutectic, LiCl--KCl eutectic, Bi-- U fuel, and BiCl/sub 3/, either separately or in various combinations. Screening tests to determine the resistance of a wide range of commercial alloys to the corrodents were performed in static and tilting-furnace capsules. Some ceramic materials were tested in static capsules. Largerscale tests of metallic materials were conducted in thermal convection loops and in a forced circulation loop. Some of the tests were conducted isothermally at 500 deg C, and others were performed under 40 to 50 deg C temperature differences at roughly the same teinperature level. On the basis of metallographic examination of exposed test tabs and chemical analyses of corrodents, it was found that the binary and ternary eutectics by themselves produced little attack on any of the materials tested. A wide variety of materials including 1020 mild steel, 2 1/4 Cr--1 Mo alloy steel, types 304 (ELC), 310, 316, 347, 430, and 446 stainless steel, 16-1 Croloy, Inconel, Hastelloy C, Inor-8, Mo, and Ta is, therefore, available for further study. Corrosion by the …

In the Sulfex process, stainless steel cladding, generally 304-L, is removed from fuel elements by dissolution in boiling four to six molar sulfuric acid. When cladding removal is complete, the decladding solution is removed is removed to waste. The exposed fuel cores are rinsed and dissolved in nitric acid.

From foreword: This report is concerned with the investigation of an indirect cycle plant which was the subject of a joint study between Sargent & Lundy and NDA.

From foreword: This report investigates the economics and performance of alternate cycles for heavy water moderated power reactors.

The following report is the final safety analysis report for the Task 2 Radioisotope Powered Thermoelectric Generator prepared by The Martin Company. It presents analyses, tests and evaluation of the operational safety criteria for the generator.

The safety aspects involved in utilizing the Task 2 radioisotope-powered thermoelectric generator in a terrestrial satellite are described. It is based upon a generalized satellite mission having a 600-day orbital lifetime. A description of the basic design of the generator is presented in order to establish the analytical model. This includes the generator design, radiocerium fuel properties, and the fuel core. The transport of the generator to the launch site is examined, including the shipping cask, shipping procedures, and shipping hazards. A description of ground handling and vehicle integration is presented including preparation for fuel transfer, transfer, mating of generators to final stage, mating final stage to booster, and auxiliary support equipment. The flight vehicle is presented to complete the analytical model. Contained in this chapter are descriptions of the booster-sustainer, final stage, propellants, and built-in safety systems. The typical missile range is examined with respect to the launch complex and range safety characteristics. The shielding of the fuel is discussed and includes both dose rates and shield thicknesses required. The bare core, shielded generator, fuel transfer operation and dose rates for accidental conditions are treated. mechanism of re-entry from the successful mission is covered. Radiocerium inventories with respect to …

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An exploratory program on subsurface coatin8s for graphite fuel elements is summarized. A number of coatings with various melting points which could be located beneath the surface of a fueled graphite sphere were investigated. Of the materials with lower melting points. nickel and a special glass compound appeared to form continuous coatings when a hot-pressing technique was employed. Several materials with high melting points. such as Ti, Cr, and MoSi/sub 2/, showed some promise, even though present equipment limitations prevented these specimens from being hot-pressed at the melting point of the coating. (W.L.H.)

This document details activities of the Fuels Preparation Department during the month of May 1960. (FI)

Results of a survey covering the safety, adequacy, and reliability of certain design and construction features pertaining to the NS Savannah nuclear power plant and related systems are presented. (auth)

Abstract: A unitized magnetic jack driven control-safety rod has been developed for the 45.5 thermal megawatt organic moderated reactor (Piqua).

A free-surface centrifugal pump, incorporating a hydraulic bearing running in sodium, was operated at the conditions required for service in the Hallam Power Reactor. After difficulties arising from inadequate shaft clearances were alleviated, the pump performed properly at a flow rate of 7200 gpm of 945 gas-cooled F sodium at 150-ft head. Results Hallam plant. (auth)

This report describes progress made in the study of the feasibility of the concept, plant arrangement studies, and considerations pertaining to partial plant attendance. It is concluded that both the indirect and direct boiling cycles are feasible. The recommended reference design is the direct cycle, in a single capsule, with the turbine, generator and pumps on a single shaft.

The thicknesses of ordinary concrete required to shield pipes in the ORNL High Flux Isotope Reactor primary water system were computed for normal operating conditions and for abnormal conditions such as a defective fuel plate or a meltdown of the fuel within the reactor. About 6 ft of concrete is required for the pipes at the outlet of the reactor, and 2 ft of concrete is required for the pipes located about 1 1/2 min downstream from the reactor vessel-. These thicknesses of concrete reduce the radiation levels to below the specified tolerances of: (a) 0.75 mr/hr during normal operation or operation with one defective fuel plate; (b) 1 r/hr immediately after the meltdown of 1% of the fuel; and (c) 1 r/hr 24 hr after a total fuel melidown. Shielding thicknesses required for other tolerances may be estimated from graphs and tables which are presented. (auth)

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In order to elucidate the phase diagram of the Zr--H system at high hydrogen contents, a pressure-compositiontemperature study of this system in the temperature range 550 to 850 deg C was carried out from ZrH to ZrH/sub 2/. From the obtained isotherms, the position of the boundary between the two-phase ( beta + delta ) region and the single phase delta region was more precisely defined (where beta designates the hydrogen stabilized high temperature zirconium phase, and delta the cubic hydride phase). The isotherm also show no evidence of a two-phase hydride region (cubic + tetragonal hydrides coexisting) in this temperature range, as has been observed at room temperature. (auth)