Report discussing an improvement program on lithium mass spectrometers.

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

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.

From introduction: The primary objective of Project 30.2 was to evaluate the protection afforded by a reinforced-concrete dual-purpose underground parking garage and personnel shelter against effects of a nuclear detonation.

Abstract: In the March-April 1962 shutdown of SM-1 Core II, the SM-28 element will be re-inserted in SM-1 Core II and an SM-1 Core I element will be removed. An SM-1 Core II europium absorber will be replaced by a Ag-Cd-In absorber, and surveillance specimens will be inserted above the core support structure. Analysis of these changes concludes that re-insertion of the SM-2B stationary element and insertion of surveillance specimens do not affect hazards potential previously defined for SM-1. Replacement of the europium absorber by the Ag-Cd-In absorber will have negligible effect on reactivity control worth of the rod. The absorber meat section is encapsulated to prevent exposure of silver alloy to the primary coolant; postulated release of silver due to a cladding defect, after 2 years irradiation in SM-1, would not cause a hazard such as to restrict access to the vapor container. Possibility of steam formation in the air gap between the absorber core and cladding, causing a cladding failure, is remote. Deformation of the absorber section sufficient to cause the rod to stick, would not impair the ability of the other rods to shut down the reactor safely.

Quarterly design and development progress report on the activities of the Army Boiling Water Reactor (ABWR) Program.

Operation and maintenance procedures for 30 megawatt heat exchanger and steam generator for sodium cooled reactor system.

From abstract: "This section, Part II, extends the comparison of two ceramic fuel systems to include the fuel cycle cost comparison in greater detail particularly with respect to fabrication and reprocessing unit costs."

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO₂ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel.

From abstract: "Describes Heat Transfer Reactor Experiment No. 3, a solid-moderated nuclear power plant with a horizontal reactor."

Report regarding ongoing research and development undertaken at Hanfdord Laboratories' physics and instruments laboratory and chemical laboratory.