Abstract: A free-surface centrifugal pump, incorporating a hydraulic bearing running in sodium, was operated at the conditions required for service in the HNPF.

Abstract: The measurement and calculation of the reactivity absorbed by Xe135 as functions of time after shutdown for the SRE are described.

Abstract: Irradiated fuel elements from the Organic Moderated Reactor Experiment (OMRE) first core loading have been examined and evaluated to determine: (1) the stability of the floating plate fuel element design, (2) the stability of the stainless steel clad UO2 - stainless steel cermet core fuel plates under irradiation and exposure to the organic coolant, (3) the extent and nature of deposits on the fuel element services, and (4) the distribution of burnup in the fuel elements.

Abstract: This report describes an advanced sodium-cooled, graphite-moderated nuclear power plant which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency.

Abstract: It has been proposed that a fuel element containing porous uranium be used for moderately high temperature applications. In an effort to determine the structural integrity of this type of fuel body, thermal cycling tests have been conducted on porous uranium.

Abstract: A portable assembly for handling and shipping OMRE fuel elements is described and details of its operation are given. Problems of heat transfer and radiation shielding are discussed, and detailed analysis are presented.

Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.

Abstract: Accurate measurement of OMRE fuel-element surface temperatures is important for evaluation of the performance of organic compound as moderator and coolant materials in nuclear reactors; a mixture of polyphenyls has been used in reactor operation to-date.

Abstract: Equipment has been developed to handle the fuel elements and control rods for the Piqua (OMR) reactor.With the handling machine, which consists of a shielded cask mounted on a gantry, a fuel element can be replaced in the core in about 27 minutes.

Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.

Abstract: Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of commercially available valves considered for use in the sodium coolant system of the Hallam Nuclear Power Facility.

Abstract: A fuel material evaluation was made by destructively examining a full-scale experimental fuel element, irradiated in the SRE to a maximum of 850 Mwd/MTU.

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

Abstract: A generalized equation of state developed by Gouq-Jen Su and Chien-Hou Chang is used for predicting P-V-T behavior of BF3 at high gas densities.

Abstract: An experimental study of valve stem freeze seals was undertaken as part of the effort to develop components for high temperature service in advanced sodium-cooled reactor systems.

Abstract: This report describes a 300,000-kwe, sodium-cooled, graphite-moderated nuclear power plant based on existing technical information.

Abstract: This report is a description of a 200 channel digital recording system used to record high temperature strain gage outputs and associated temperatures.

Abstract: A device to detect the presence of hydrogen in sodium has been developed. Such a device, installed in a sodium heated steam generator, would signal the presence of water in the sodium resulting from a leak in the sodium-water barrier.

Introduction: This is the fourth report in a series of reports dealing in particular with a reactor safety device of a double diaphragm type.

Abstract: The means to prevent the recurrence of tetralin leakage into the SRE sodium systems are discussed. Included is a description of the redesign of system components to utilize alternate coolants such as nitrogen, air, and NaK.

Abstract: Because of the outstanding heat transfer efficiency of sodium, it is necessary in sodium-cooled reactors to consider and attempt to prevent the occurrence of adverse stresses as a result of thermal transients in the system.

Abstract: A 3000 kw (thermal) bayonet duplex tube model steam generator was performance-tested in a liquid metal test loop at MSA Research Corporation, Callery, Pennsylvania, under the cognizance of Atomics International.

Abstract: A multigroup P1 approximation for hydrogen scattered neutrons has been developed and applied to the study of neutron flux distributions in titanium hydride and steel shield systems.

Abstract: This report describes the fabrication of elements for the OMRE third core loading.