Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).

Abstract: Design charts and systematic design forms are presented for simplified calculations to check the number of convolutions and thickness required to limit the deflection and pressure stress range in three types of bellows.

A report regarding irradiation behavior of uranium monocarbide fuel experiments NAA 81-3 and AI 3-12

From abstract: The fuel element assembly was successfully irradiated in the SRE to a maximum burnup of 5300 Mwd/MTU, at a peak fission rate of approximately 1.5 x 10E13 fissions/cm3-sec and a maximum central temperature near 1200F.

Abstract: The data presented herein, in the form of graphs, can be used to obtain the value of this energy.

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

Progress report for the Piqua Reactor Operations Analysis Program describing observations and analyses at the Piqua Nuclear Power Facility (PNPF). The program goals are to monitor operations and collect data in order to ensure that the plant's operation is safe, to improve design and performance, to evaluate the performance and lifetime of the plant's components and systems, to evaluate plant safety and safeguards, and to disseminate all information to the scientific community.

Progress report for the Piqua Reactor Operations Analysis Program describing observations and analyses at the Piqua Nuclear Power Facility (PNPF). The program goals are to monitor operations and collect data in order to ensure that the plant's operation is safe, to improve design and performance, to evaluate the performance and lifetime of the plant's components and systems, to evaluate plant safety and safeguards, and to disseminate all information to the scientific community.

Abstract: The results indicated that increased reliability can be obtained by using thermocouples made with insulation of increased density and/or a low thermal expansion sheath.

Abstract: FAIM is a general multigroup, one-dimensional diffusion equation code programmed in FORTRAN language for the IBM 7090 computer.

Abstract: An investigation of the metallurgical aspects of the SRE fuel element episode, that occurred July 26, 1959, has been completed.

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: This report presents the results of plant optimization studies and cost estimates of the reference design for a natural uranium, graphite moderated, gas-cooled reactor and power plant which was described in NAA-SR-1833.

Abstract: Presented herein is the preliminary design of a natural uranium, graphite moderated, CO2-cooled reactor and power plant similar to, but larger than, the British Calder Hall plant, with a net electrical output of 130 MWE.

Abstract: The experimental application of centrifugal clarification, precoat filtration, conventional filtration, and adsorption to the removal of impurities from a bypass stream of irradiated reactor coolant at the Organic Moderated Reactor Experiment is described and evaluated.

"Scale-up experiments on high temperature fuel recovery processes have included the dummy run phase on the handling of 1-kologram samples of molten, non-irradiated uranium in the hot cell. The next step involves the use of spent X-10 fuel slugs. Small scale experiments with X-10 uranium on the extaction of Pu with Mg show that as much as 80 per cent of the Pu can be removed in pone pass. Treatment of uranium with fused fluorides can remove at least 90 per cent of the Pu in one pass. Oxide scavenging with ZrO2 is very effective in removing rare earths.:

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.

A preliminary study has been made of a solid homogeneous reactor for stationary power plant application. The core consists of graphite spheres impregnated with uranium and thorium, and the coolant is bismuth. This concept possible offers advantages over other solid fuel reactor systems with respect to simplification of core structure, fuel fabrication and fuel handling, and reduction of fuel inventory external to the reactor. From the results of this preliminary study, it appears that the potential cost of electric power from this reactor is competitive with that from other reactor systems which have been proposed for the same application. The Po210 produced in the coolant presents a decontamination problem, but is also possibly a valuable by-producgt.

Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.

"This report describes second-cycle postirradiation examination and AIROX reprocessing-refabricating of uranium dioxide irradiated to an additional 10,000 Mwd/MTU burnup."

Introduction: QUICKIE is a computer program designed to solve the multigroup neutron slowing down and thermalization equations without consideration of spatial dimensions.

Abstract: This report describes in detail two designs of a nominal 6-kwe Nuclear Power Plant (NPP), one using thermoelectrics for power conversion and the other using the Mercury-Rankine cycle NPP.

Abstract: this report discusses and summarizes all stress analysis done on the SNAP 10A system; it also mentions many of the structural tests which were accomplished.

Abstract: A series of clean critical experiments was performed in the SGR critical facility utilizing 2 wt % enriched, uranium metal, hollow cylinder, fuel elements, in AGOT graphite moderator.