Increased interest in the civilian, space, and military applications of isotopic power has prompted a study by which the Chemical Processing Department can provide the interim strontium-90 requirements on a schedule considerably accelerated from that previously proposed. This document summarizes a study of the technical and operational feasibility and hazards involved in: recovery of semi-refined megacurie quantities of Sr{sup 90} from current Purex waste; lag storage of the Sr{sup 90} fraction in the 244 CR Process Vault for Sr{sup 89} decay; and subsequent reconcentration and refinement to a bulk product suitable for isolation and packaging by Hanford Laboratories.

The objective of the production test detailed in this report is to compare the dimensional stability characteristics of fuel elements with alloyed low hydrogen dingot cares and standard fuel elements with ingot cores. The basic measurements of dimensional stability are the average warp and the tube-filling capacity values of the fuel elements.

The objective of the production test detailed in this report is to evaluate the effect of a 36 mil Zircaloy-2 outer jacket on the behavior of heavy walled tubular elements during high temperature irradiation. Zircaloy-2 jacketed unalloyed natural uranium fuel elements nominally 1.43 inch OD, 0.52 inch ID, with 36 mil jackets on the outer surfaces will be irradiated in the KER Loops to an exposure not greater than 3500 WD/T. Several of the fuel element failures that have occurred during testing in the KER Loops have apparently resulted from localized jacket thinning on the external surfaces of high exposure elements. One of the methods proposed to prevent this failure mechanism was to increase the Zircaloy-2 jacket thickness on the fuel element outer surface. The fuel elements authorized by this production test have been prepared with nominally 36 mil external jackets. The behavior of these elements will be compared to the behavior of similar elements with 20 mil jackets currently being irradiated in KER-3.

The objective of this production test is to evaluate the behavior of brazed fuel element end closures prepared with a 12 w/o Iron, 4 w/o Beryllium, 84 w/o Zirconium braze alloy during high temperature irradiation in the KER Loops.

The DR-1 Loop, located in the C test hole of the DR Reactor, provides a high temperature, recirculating gas-cooled facility for the irradiation of experimental fuel elements. The loop is being utilized currently by General Atomic, a Division of General Dynamics, to evaluate fuel elements in support of their work on the Maritime Gas-Cooled Reactor Program, a program which is directed at the development of a ship propulsion unit consisting of a gas-cooled reactor driving a closed cycle gas turbine. The loop irradiations for this program require that the experimental fuel elements be maintained at specific test conditions. It is also necessary that all of the loop components be kept within certain operating limits. Therefore, the power generation rate of each experimental fuel element must be evaluated and established as accurately as possible prior to insertion in the loop. One method of establishing the enrichment required to obtain a required heat generation rate in an experimental element is to irradiate a nuclear mock-up of the assembly in the Hanford Test Reactor to determine the relative neutron density within the assembly and the reactor. This report presents the results of such irradiations using the MGCR-III mock-up.

Canning bath Al-Si, supplied from offsite vendors and reclaimed lathe turnings in the 313 building, is used in the production of I & E fuel elements. A study was made of the effect of increasing the Ni content to over 0.5% in the canning baths, in order that all of the X-8001 scrap could be reclaimed. Effect on bond quality, weld integrity, and canning bath operation was studied. Based on adverse weld quality, slight loss in reactivity, and potential for furnace channel plugging, it is recommended that the present Ni specification of 0.5% maximum remain unchanged.

Experiments have been conducted in the Hydraulics Laboratory, at the request of IPD`s Mechanical Development-A Operation, to determine the energy losses of various enlarged outlet fitting combinations. These experiments were conducted an steady state runs and allow the determination of the normal operating point (flow rate) of a reactor process channel under selected conditions of front header pressure and fuel charge. No attempt is made to make a mechanical or economic evaluation of the particular fitting combinations, although observations were noted which might bear on this evaluation. It is very important for the reader to bear in mind that changing outlet fittings will definitely affect the reactor tube power limits and outlet vater temperature limits. The size of the outlet fittings largely determines the present outlet temperature limits of the old reactors. The flow characteristics of these present fittings cause some degree of pressurization to suppress boiling on the fuel charge and also cause dual Panellit trip protection for certain flow changes and for power surges. Enlargement of the outlet fittings may actually reduce the allowable outlet coolant temperature limits. Since these effects cannot be determined on the apparatus used in these experiments, a complete discussion of this point is …