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Abstract: The fast and thermal neutron leakage from the 30 megawatt SGR-P4 reactor has been studied by three independent methods.

A report regarding the irradiation behavior of Unalloyed hypostoichiometric uranium carbide experiment AI 3-11 and review

Abstract: Under the Power Expansion Program (PEP), modifications have been made to the Sodium Reactor Experiment (SRE) facility to improve plant reliability and permit an increase in power to 30 Mwt, with a reactor coolant outlet temperature up to 1200°F.

From abstract: The Madelung constant and the inverse twelfth power repulsion factor have been calculated for the wurtzite structure for wide ranges of the crystal parameters and u.

Abstract: An intuitive approach to the understanding of crystal structures is presented in terms of the concept of the closest packing of spheres.

"An intuitive approach to the understanding of crystal structures is presented in terms of the concept of the closest packing of spheres. The qualitative features of the concept are sorted out and correlated by successively treating single, double, triple, and multiple layered arrays of closest packed spheres" (p. ix).

The design of a waste-gas treatment system for organic moderated reactors requires a knowledge of reactor waste-gas composition, generation rate, and radioactivity. To obtain data on these variables a continuous stream analyzer was constructed to analyze the waste gas from the Organic Moderated Reactor Experiment (OMRE).

Abstract: This report presents a reference design of an organic moderated and cooled reactor for the propulsion of a 38,000-dwt tanker.

Introduction: This report has been prepared to document the extensive analytical work required to design the control system for the Sodium Components Test Installation (SCTI), constructed for the Atomic Energy Commission at the Atomics International field laboratory.

Abstract: One hundred gram quantities of uranium dioxide, irradiated to burnup as great as 21,000 Mwd/MTU and previously reprocessed by AIROX (Atomics International Reduction Oxidation), were refabricated into high density pellets and encapsulated for re-irradiation.

This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.

From abstract: A series of three multi-compartmented experiments was designed to investigate the irradiation behavior uranium carbide fuels, as a function of composition, temperature, and burn-up; and the mechanical properties of Type 304 stainless steel, as a function of temperature and integrated thermal neutron flux.

Abstract: An analytical and experimental effort toward a better understanding of the mechanisms involved in nuclear power reactor loss-of-pressure accidents is presented.

Introduction: This memo gives a general description of the components and equipment affecting the control of the SRE, and the equipment associated with all reactor services.

Abstract: Cast and Wrought specimens and restrained wrought specimens of unalloyed uranium were irradiated in the Materials Testing Reactor, as the first in a series of experiments to develop fuel materials for sodium cooled reactors.

"Heat production resulting from the absorption of gamma ray photons in thermal shields and the leakage of neutrons and photons from ferritic thermal shields are investigated. The gamma rays considered arise from three types of reactor radiation -- thermal neutrons, fast neutrons, and core and reflector gammas. The energy spectra of the fast neutron leakage and absorption have been investigated in some detail because of the significant contribution of fast neutrons to the heating of the concrete biological shield."

Abstract: This report presents the results of the structural tests which were performed on the S10A-PSM-1A system to qualify the basic structural design for the anticipated launch environment.

From abstract: A conceptual design study has been performed for a sodium cooled, graphite moderated, thermal power-breeder reactor utilizing the Thorium-Uranium 233 breeding cycle. Several aspects of the design of the system are considered but no attempt has been made to supply all the details. It appears that the design presented is feasible and will allow the production of economic power as well as full utilization of thorium resources.

Introduction: A series of critical and subcritical experiments have been carried out as part of the Advanced Sodium Graphite Reactor (ASGR) Critical Experiments program to provide physics information which would be directly applicable to the AGSR concept.

Introduction: The purpose of this experiment is to determine the loss of electrical power output due to operating many unequally-heated cesium diodes, connected in series and parallel circuits.

Abstract: A magnetic-jack control-safety rod is under development for the 45.5 thermal megawatt Organic Moderated Reactor. The rod is "unitized," i.e., the poison element, drive, position indicator, and shock absorber are contained in a compact assembly which is inserted in a regular fuel channel opening in the core. Tests to develop components capable of operating under these conditions are described and results are reported.

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

"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."