Technology of the Sodium Graphite Reactor. Reactivity calculations were made to study the application of steadystate plutonium feedback techniques to the use of diffusion plant tails for reactor fuel feed material. The performance and design of a twin core SGR power plant are given. Thermal neutron flux distribution measurements are reported for a six-rod fuel cluster and for a hollow uranium rod. A power cost calculation was made for a 1000-Mw SGR Th-U/sup 233/ breeder reactor which starts up on Th--U/sup 235/ alloy. Calculations were made on neutron leakage through the SRE shield. Research on reactor fuel elements and reactor materials are described. Corrosion and irradiation damage data at 5 x 10/sup 7/ r dose (150 deg F) on toluene as the SRE shield coolant indicate that the radioinduced corrosion of Fe, Al, and Cu in the SRE shield should be negligible. Preliminary results are summarized for l-Mev electron ir radiation studies of terphenyls at 400 to 450 deg C. Sodium Reactor Experiment. Progress is reported for various portions of the SRE project: reactor design and evaluation, fuel elements, moderator, reflector, structure, reactor cooling and heat transfer, instrumentation and control, shielding, and reactor services. (D.L.C.)

Project CG-588 authorized the design and construction of dissolver and waste neutralizer off-gas scrubbers to remove the ammonia given off during coating removal and waste neutralization steps of the Redox operation. It has always been recognized that the nitrogen oxides in the dissolver off-gases, resulting from the dissolution of bare uranium slugs, could also be absorbed in water under proper conditions to give re-useable nitric acid. Consequently it appeared feasible to provide facilities which would combine these ammonia removal and nitric acid recovery operations. The purpose of this report is to present a scope design for the economical recovery of nitric acid from the Redox dissolver off-gases in addition to the removal of ammonia. It is recognized that acceptance of this scope for project execution would make unnecessary the ammonia scrubbers for the dissolver off-gases of Project CG-588. 8 refs.

The minimum shutdown time, to permit accessibility, for the Intermediate Power Reactor is estimated to be 38 hours. In case the reactor were shutdown following each rupture this long shutdown period would have serious disadvantages. The desirability of being able to make firm power commitments (independent of slug ruptures) has led to a study of the possibility of continuous operation following a rupture. There is evidence to indicate that, at the proposed water temperature, the rate of corrosion of uranium may be so high that at least a major portion of the rupture products may have entered the system before the reactor can be shutdown. A pushout of the affected column would then be a pushout of only those slugs which are still intact and the problem would still remain of removing the rupture products from the system. The first portion of this report is concerned with the rate of corrosion of a slug following rupture and the possible limitations to the principle of non-shutdown operation. These limitations include a flow stoppage by the ruptured can, undue increase in gamma activity, increased corrosion by the rupture products, and adherence of rupture products to the piping. The latter portion of the …

The purpose the production test described in this document was to evaluate by pile irradiation the stability and resistance to rupture of powdered metal compacts.

The purpose of this document is to present a capacity study of the existing UO{sub 3} Plant Nitric Oxide Absorption System.

The objective of the experimental studies reported herein was to demonstrate the laboratory-developed mercury-catalyzed dissolving procedure on a pilot-plant scale by conducting a series of test dissolutions of aluminum-jacketed, unirradiated uranium in the 150-gallon downdraft type semiworks dissolver in 321 Building. It was desired to establish: (1) the effect of the larger-scale operation on time cycle; (2) dissolution rates of aluminum and uranium, (3) controllability of the reactions, and (4) the concentration of mercury catalyst required.

From April, 1951 to Aug, 1954, New York Shipbuilding Corp. carried out a subcontract with E.I. du Pont de Nemours & Company that was without parallel in the shipyard`s history. The work, designated the NYX Project for reasons of security, was vital to the operations of the Savannah River Plant, Aiken, S.C., which was then being designed and constructed by du Pont for the Atomic Energy Commission. It consisted of three broad parts: developmental and experimental work; fabrication and testing of a prototype unit; fabrication of production units. Five production units were ultimately built, one of them converted from the prototype. All were fabricated from stainless steel, and involved welding techniques, control of thermal distortion and tolerances never previously attempted on assemblies of comparable size. Du Pont`s technical experience and the background of New York Ship in heavy construction, particularly in the fabrication of naval gun turrets, were combined from the outset to resolve the difficult fabrication problems that occurred almost daily. Representatives of both companies worked together as a team in the shops and at supervisory levels to an unprecedented extent. The report is intended primarily to summarize New York Ship`s part in the project, but also includes some …

Free-flight drop vehicle tests have been made to investigate devices for measuring ambient pressure in the vicinity of a high-fineness-ratio weapon shape throughout the transonic speed range. Various types of nose probes and trailing probes were tested.

This report presents results radiation monitoring from the separations section department for December 1953. Information is presented concerning radioactive effluent releases, personnel contamination, and general radiation monitoring from the facilities.

The role of ionization in the phenomenon of" radiation annealing" of graphite has been studied by using a I-MeV electron beam. Changes in the C-axis of a sample with a Hanford irradiation of 460 Mwd/ct were studied. Two thermal anneals of 4 hours each at 350{degree}C proved sufficient to complete the thermal annealing at this temperature. The samples were then irradiated for 7 {1I2} hrs at a temperature of 340 {degree}c. The samples received an irradiation of 47 microampere-hours, equivalent in ionization to an exposure of 200 Mwd/ct in a Hanford reactor. No changes were noted as a result of the electron bombardment. It is concluded that the ionization is not of major importance in radiation annealing.

The tables and graphs of this publication are essentially an expansion of the 'Range-Energy Curves' of Aron, Hoffman, and Williams (AECU-663). The original work of Aron et al is extended to include pions, protons, deuterons, tritons, and He{sup 3} particles in selected absorber elements and in some compounds that are useful either as target materials or as scintillating materials. This extension is an outgrowth of needs that arose during the course of certain experiments that have been carried on for the past several years at the University of California Radiation Laboratory at Berkeley. Some of the work was done in a random fashion as needed chiefly for pion production experiments with scintillation counters. It seemed worthwhile to unify and complete this work for its utilitarian value to experimenters in high-energy physics, who have found similar needs.

This report addresses the basic research and particle accelerators.

The precise relationship required between the frequency of the accelerating voltage and the magnetic field guiding particles accelerated in a high-energy proton synchrotron presents a challenging design problem ot the electronic engineer. In the University of California Bevatron, it has been necessary to develop a number of circuits whose performance follows a prescribed function to within better than one-tenth of one per cent. Control and monitoring of the acceleration cycle is carried on at a control area suitably shielded from high-energy radiation.

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A preliminary evaluation of the hazards associated with a 20-Mw boiling reactor for the purpose of determining site requirements is presented. The Boiling Experimental Reactor design, safety features, and performance are given and the surroundings of the site at Argonne National Laboratory are described. (T.R.H.)

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