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.

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.)

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)

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.

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.

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)

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.

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)

A study to improve the elevated-temperature strength of niobium by solloving has resulted not only- in greatly improved strengths at 1200 and 1470 deg F but also in the development of improved fabrication techniques for these alloys. The most important step in the fabric:ition procedure of niobium and niobium-base allows is the initial breakdown of the cast structure. The cast structure of 1.84 wt. 4 chromium, 3.21 wt.% chromium. 4.33 wt. ' zirconium, and 20.5 wt.% titanium-4.28 wt. = chromium allovs and unalloyed niobium was broken known by- forging ingots (protected from oxidation by molybdenum ciins) at 2550 deg F and rolling at 800 deg F. After the initiai breakdown of the cast structure, the alloy-s were coid roiied to a total of 95 per cent reduction with no difficulty .A second fabrication technique was employed for a second set of alloys. Unalloyed niobium and 1.29 wt. % chromium, 2.74 wt. 3 zirconium, 4.5 wt.% molybdenum, and 10 wt. % titanium-3 wt.% chromium alloys were forged and rolled at 1000 deg F to break down the cast structure and then cold rolled to 0.030-in. sheet. the sheet obtained by this technique showed moderate edge cracking. Tensite tests on the coid-worked …

The Criteria for Speed of Control must be met for all cases being considered in the Expansion Studies. These Criteria are dependent upon the ratio Volume of water/volume of uranium in the process channel, the reactor power level, and the rate at which the vertical rods are released and inserted into the reactor. For those cases wherein the graphite channel is overbored by 0.200 inch., the criteria can generally be met for the power levels under consideration since the ratio of Vw/Vu is maintained adequately small by the use of large diameter fuel elements. If the graphite is not overbored, high pressure drop across the fuel element section must be maintained in order to maintain small values of Vw/Vu, which results in a substantial increase in the front header pressure. A method is developed in this document which ties together and defines the hydraulic and physics characteristics of the fuel-process channel geometry necessary to satisfy the Speed of Control Criteria over a range of feasible operating conditions. This study assumes no graphite overboring and employs throughout a smooth bore zircaloy process tube having an inside diameter of 1.650 inches. The fuel elements are self-supported., and the inner hole size was …

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.

This document details activities of the Fuels Preparation Department during the month of January 1960. (FI)

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 …

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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)

This document details activities of the irradiation processing department during the month of December, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

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.

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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 study proposes improvements in the process water piping adjacent to the front and rear faces of these reactors. This report covers the external piping of the reactors from the incoming valve pit to the inlets of the outgoing retention basins.

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.

The objectives of this production test detailed in this report is to evaluate the behavior during irradiation of tubular fuel elements with hot-headed end closures. With natural uranium twenty-inch tube- and-tube elements will be irradiated to a goal of 2500 MWD/T in the KER loops. The inner tubes will be closed by the hot-heading technique and the outer tubes will have normal welded closures.

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.