Fuel element corrosion studies designed to and in selecting and evaluating SM-2 fuel element welding techniques are discussed. Tests on type 347 ss plate type fuel elements welded by the selected tungsten inert gas technique showed good corrosion integrity of specimens under a variety of conditions including a 500-hr test under simulated SM-2 conditions of flow and coolant chemistry. (auth)

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Experimental data on plutonium yield and U{sup 235} burnout are now available on the E metal portion of three central zone striped E-N columns and one fringe blanket E column. These data and an overall E-N load conversion ratio based on experimental data are now reported.

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Production of Pu nitrate from separations plant during this month was below forecast. UO{sub 3} production, shipments met schedules; so did unfabricated Pu shipments. Purex plant was stopped to replace the final waste concentrator (F-11) (leak). Two Pu-U partition failures were attributed to foreign organic material in the nitric acid; the Pu product was kept within specifications by adding NaF and ANN to ion exchange feed stream. A Np recovery run was started in Redox, and dissolution was started of 12 special 2-ton test batches of normal U fuel elements, irradiated to provide information on Pu formation rates. The damaged B-2 E-metal dissolver was replaced with a conventional dissolver. Test of a new sieve plate cartridge in Recuplex H-1 extraction column was stopped. A new semi-continuous product concentrator-stripper was made to replace Recuplex batch concentrator. Conversion of Purex prototype anion exchange to a manufacturing unit is nearly complete. Design was completed on the new Redox E-metal dissolver. Process feed was introduced into RMC button line and 3 buttons made. Project proposal for NPF reprocessing was revised.

Report documenting ongoing research and developments at the Oak Ridge National Laboratory's Gas-Cooled Reactor Project. Design Investigations: The effects on the power distribuestablished. A mathematical model was developed for studying shifting of the coolant stream as it moves along a rod in order to predict the temperatures of the parallel streams as they progress through the reactor. A fuelelement life code developed for computing the internal temperature structure, the amount of fission gas released, the internal pressure, the cladding strain when the internal pressure exceeds the coolant pressure, and the creep damage was used for comparing top-loading and inventedloading fuel programs for the EGCR. A statistical method was developed for estimating the probability that the hot spot on the EGCR fuel element will exceed a given temperature. A method of cooling the EGCR control rods was developed that will minimize diversion of coolant flow through leakage paths between graphite blocks. A preliminary design of a control rod cooled by this method was developed. Means for reducing the thermal stresses in the top head nozzles of the EGCR pressure vessel were studied. The stresses in the graphite sleeves of the EGCR fuel elements were calculated, and the maximum stress was found …

Studies on the use of coagulant aids in the lime-soda treatment of large- volume, low-level radioactive liquid waste revealed that a combination of Hagan Aids No. 50 and No. 18 gave fairly good results under most conditions. The effects of feed solution concentrations, mode and point of addition, and water temperature were studied. (C.J.G.)

An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)

Use of X-8001 Al alloy as cladding for Hanford reactors was initiated because of superior (laboratory) resistance to intergranular corrosion over that of C-64 alloy. However, since severe pitting attack was observed intermittently, an evaluation was carried out on X-8001 alloy fuel element cladding.

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An investigation was made to determine the structural behavior of selected components of the Oak Ridge Research Reactor for increased power level conditions. It was found that a reactor cooling water outlet temperature of 150 deg F will cause severe plastic strain cycling in the aluminum housings for the large test facilities. Increasing the reactor cooling water flow rate of 21,000 gpm will cause plastic deformations in certain reaons of the core box. These latter deformations can be tolerated, but the full implications asscciated with any change in pressure differential must be understood before adopting the above flow rate. (auth)

In an investigation of the detailed temperature structure of the seven- element cluster of cylinders surrounded by a sleeve which comprise the fuel assembly for the EGCR, the radiative interchange of heat between the rods and sleeve was evaluated. A procedure advocated by Hottel was used to determine the view factors for gray enclosures taking into account the multiplicity of reflections, absorptions, and secondary radiations. (auth)

As a result of atomic shelter tests and field experiments condueted over the past nine years, it has been conclusively shown that shelters provide the only promising means of civilian protection in the event of a nuclear war. Design details are presented for a group shelter to accommodate 100 persons of all age groups and both sexes. The shelter structure is a multiplate corrugated- steel arch set on a concrete slab with end walls of bridge plate sheathing. The entire structure is covered with a minimum of 3 feet of earth. The shelter combines outstanding protection against radioactive fall-out with good protection against blast and thermal radiation. Drawings are included. General operating procedures are outlined. (C.H.)

Seven water leaks during the past year because of internal corrosion of second generation process tubes have emphasized the need for additional corrosion information. Numerous out-of-pile and probolog examinations have been made to determine the nature of the corrosion, and mixer fuel elements currently are charged in central zone tubes in order to reduce the corrosion rates. This document presents the data obtained to date as a general description of the tube condition at the time when the mixer elements were adopted and outlines future action necessary to determine when the second generation tubes must be replaced because of internal corrosion.

This report presents a technical and economic study of the lead-dip canning process compared to two types of solid state diffusion bonding processes (hot-press die and gas pressure bonding) for Hanford production fuel elements (exclusive of NPR). It is concluded that either bonding process has potential advantages over the lead-dip process. The gas pressure bonding process has advantages over the hot-press die process.

This report presents provisional specifications for the modifications and additions to the Lead Dip Canning Process for Heat Treated Uranium for the production test fabrication of projection fuel elements by resistance spot welding.

1960. 6p. OTS. Depth measurements were performed in the Screenhouse intake channel and in the Ohio River in the vicinity of the Shippingport PWR Screenhouse. The need for dredging was established. (C.J.G.)

A preliminary analysis conducted on various types of bearings indicated that hydrodynamic type journal and thrust bearings lubricated with a portion of the mercury from the condensate return pump would best suit the SNAP II requirements. Experimental rssults confirmed the bearing design approach. Stable bearing operation was obtained at speeds in excess of the 40,000 rpm design objective with simulated loads of 1 to 10 g in the radial direction, and 0 to 2 g in the axial direction. Total power consumption of the bearing system is approximately 550 watts at the design speed. (auth)

The effect of isolation and forced draining on the radiation level of the 1AC coolant purification hairpin loop was investigated. The activity in the horizontal and vertical legs (U-bend) of the loop piping was determined by a radiochemical analysis of the crud deposits. (C.J.G.)

This report is an interpretation of data (flow, pressure, temperatures within a process tube during events affecting single tubes) as applied to the most severe (rapid) K-reactor transients which are credible. Analyses indicate that no fuel channel burnout will result from a BPA power loss to the process pumps.

This document details activities of the irradiation processing department at the Hanford Reservation during the month of February, 1960. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operation; Manufacturing Operation; Facilities Engineering Operation; Relations Practices Operation; Financial Operation; and NPR operation.

The proposed increased water capacity for 100-K plants will increase the electrical load to be supplied. The load study showed that the capacity of the existing 13.8 kV system is adequate to carry the increased loads proposed for Project CG-775, while for the 5 kV system, an expanded power system is proposed. Likewise, the voltage regulation on the kV system bus will be excessive, and voltage regulators should be added.

Production of Pu nitrate, UO{sub 3}, and unfabricated Pu metal met schedules. Decontamination performance of Purex process continued below standard. The cerium-144 cask is being redesigned. A ``powered ferret``, for driving a scintillation counter through a conduit to monitor ground activity beneath waste storage tanks, is being designed.

Approximately 100 ternary and quaternary spongezirconium alloys were screened for structural and cladding applications in a natural-uranium-fueled heavy-watermolerated power reactor. The alloy additions studied included2 to 4 wt.% Sn, 0.5 to 2 wt.% Mo, and 1 to 3 wt.% Nb. The effect of 0.1 wt.% Fe and 0.05 wt.% Ni additions to the experimental alloys was evaluated. All compositions were are melted, rolled at 850 ction prod- C from a helium- atmosphere furnace, vacuum annealed 4 hr at 700 ction prod- C, and furnace cooled. Room- and elevated-temperature hardness measurements were used to estimate the tensile strengths of the alloys, while corrosion resistance was evaluated by 1000-hr exposures to static 300 ction prod- C water. (auth)