Abstract: "Tungsten-rich alloys, developed for use at temperatures up to 2000F, exhibit ductility, fabricability and joinability not found in commercially-available materials. An envelope type of microstructure was produced in compositions containing at least 90 wt% tungsten by liquid-phase sintering of cold-pressed powders in hydrogen. At room temperature the alloys could be rolled extensively, and tensile elongations up to 25% were noted. Strength properties of a W-Ni-Fe base were improved by small quaternary additions. The ultimate tensile strength of a 90W-4.8Ni-3.2Fe-2Ru alloy was 46,700 psi at 2000F, compared to 30,000 - 35,000 psi for unalloyed tungsten or W-Ni-Fe; the 100-hour stress-rupture strength at 1600F was 15,000 psi. Excellent joints were produced by spot welding and localized induction heating. The oxidation resistance of unprotected 90 wt% tungsten compositions was not significantly affected by alloying."

The following document is a note describing the process of building a nuclear density gauge and its improved features in comparison to previously reported density gauges.

This document discusses waste disposal criteria were established on the basis that the occurrence of river flow rates which were 72 cents of normal would not cause the effects of waste disposal to exceed limits. Since even the base case exceeds the criterion for the average body burden of phosphorus-32, provision to reduce the output of this radioisotope must be included in any expansion program. Provision to reduce the output of other radioisotopes will be required for most cases where the bulk outlet temperature limit is 105{degrees} or higher. For reactor flow rates exceeding 100,000 gpm it may be necessary to reduce sodium dichromate concentrations as low as 1.5 ppM during periods of low river flow. Heat output was discussed but no limit was set.

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If material received under this program is deemed acceptable for reactor charging, this test will authorize irradiation of fourteen monitor columns which shall be discharged at 200, 400 and 800 MWD/T to observe the performance of this material. This test shall run concurrently with irradiation of quantities up to 60 tons/month to be authorized separately.

Laboratory data of flow rate versus header pressure were obtained for various conditions of nozzle cap removal from C, K and BDF single tube mockups of central zone tube assemblies using I&E slug charges. The data are presented. Suggestions are made for applying the data to DR and H reactors. In general, the effect of a pre-inserted support charge on the flow rate is small, especially with the front nozzle cap on. It should be noted that pre-insertion of an entire (117 inch long) support charge and subsequent front cap replacement is impossible in either a BDF tube with 34 fuel elements or a C tube with 32 fuel elements simply from a length standpoint.

This report records the basis underlying the decision already reached to recommend a change of process tube fittings at B, D, F, DR and H reactors, and also proposes for consideration a change in the devices for monitoring effluent water temperature.

Purpose of this study was to determine the {sup 237}Np input to the Purex process. The {sup 237}Np content of 639 g Pu/ton U irradiated fuel was found to be 1.78 {plus_minus} .09 g/ton of uranium at the 95% confidence level. Standard recovery for the chemical method was 96.7%, 98.0% for the sampling.

A report discussing "an explosive ferroelectric power supply and the experimental approach followed in determining its operating characteristics."

During the course of process engineering work at the K Reactors, a wide variety of problems have been encountered where hydraulic information was required. This information was, obtained from various documents, from calcuations, and from actual tests run specifically to obtain the information. In order to simplify work of this nature in the future and to present a convenient means of obtaining information, the data so collected are summarized and graphed in this document.

This document details the activities of the Fuels Preparation Department during the month of October 1959. (FI)

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It is the objective of this study to investigate preliminary expansion program requirements for process water, as supplied by 190 Building equipment; from the point of view of practical pumping, flywheel and pump suction head requirements. These requirements are to be determined at this time in such a form and accuracy as to be useful in refined estimating for budget study purposes. In order to obtain the objectives of this study at this time it has been decided to consider five different conditions of process water flow to a reactor. These conditions are flow to the reactor under summer conditions of operation in gallons per minute with a corresponding top of riser pressure in pounds per square inch: 85,000 gal/min, 580 psi; 100,000 gal/min, 480 psi; 130,000 gal/min, 280 psi; 150,000 gal/min, 280 psi; and 150,000 gal/min, 150 psi.

This monthly document details activities of the Hanford Operations Office during the month of October 1959. (FI)

On the 0000--0800 shift, September 30, 1959, the delayed neutron monitor on KER Loop 3 gave a high coolant activity signal. Events occurred rapidly that tended to substantiate the possibility of a fuel element failure in this loop. KE reactor was shut down immediately thereafter. This report is being written to summarize the sequence of events prior to and those following the KE reactor scram and to discuss the results and significance of data from analyses on coolant and coupon samples taken from KER Loop 3.

Pu output from separations plant was less than scheduled, but year-to- date production exceeded commitment by 4%. The Palm recovery run in Purex was the most successful to date. UO{sub 3} production and shipments met schedule. Purex had two pump failures. When Purex 1WW was centrifuged and treated to recover Ce, most of it remained in the centrifugate; only 14% was recovered. The prototype Pu ozonator in Redox performed well. Test runs on an acid precycle flowsheet and a proposed internal recycle scheme for Palm recovery were initiated in Redox. Recuplex had a change in solvent extraction feed preparation, and an installation of a safe-geometry bottom section on the stripping column. Storage of Purex 1WW wastes was discussed in a meeting. Conversion of Rexuplex to a manufacturing facility was completed. Cost estimates were developed for several alternative Palmolive processing schemes. Process flow diagrams were completed for Sulfex decladding of Yankee elements and Zirflex decladding of Dresden elements.

This report discusses two of the direct cast core I & E fuel elements irradiated in C Reactor under PT-IP-93 which were received at the Radiometallurgy Laboratory for examination during July 1959. One of the elements was badly warped and distorted while the other of the same series had remained relatively unchanged. Both pieces had received an average exposure of 689 MWD/T. Warp and diameter measurements show that the one element had warped approximately .040 in. and had become distorted in two longitudinal planes at the male end. The distortion planes are at right angles which gives the transverse section a square a appearance. A comparison of the grain size from three positions along the warped element with a section cut from the unwarped element revealed a greater number of large grains in the zones of highest distortion. The maximum size grains, were in the order of 2.5 mm -- 3.8 mm vs normal .2 mm. Also prevalent in all sections examined vas a columnar type grain structure around the outside periphery. It is evident that the pre-irradiation heat treatment failed to refine the grain size sufficiently to prevent nonuniform growth of this element during irradiation. A corrosion pit was …

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

Introduction: A low-capacity, low-decontamination plant can be built, as part of a power reactor complex, to avoid long distance transfer of fuel to a high-capacity aqueous processing plant. Activation of such a complex, with the processing plant adjacent to the reactor it serves, could decrease the cost of the integrated fuel cycle. The study of this concept is a major objective of the Processing Refabrication Experiment (PRE).

Experiments were conducted to determine the relative flow characteristics of J-pigtail adapters supplied by Lampson Products Co., Seattle.

This report provides a summary of the Dana heavy water contract as of November 1958.

Five eighteen-inch enriched tube-and tube-elements will be irradiated in the KER Loops in pH 10 water to an exposure of 3500 MWD/T, (to see if a pitting corrosion problem exists under NPR conditions).

A metallurgical bond can be effected between a Zircaloy-2 jacket and a thin walled uranium tube by treatment in a high temperature, high pressure gas autoclave. Bonding is independent of core history, provided that all surfaces are free from contamination. Gas pressure bonding causes the fuel element jacket to conform to the core, which promotes bonding on all surfaces including the ends. Grain growth occurred in some instances between the jacket and end cap interface. A three phase system (alpha, delta and epsilon) was identified across the interface at the bonding conditions of 845 C (1500 F), 10,000 psi and a time of four hours. Four structures were observed; these were epsilon, epsilon plus delta, delta, and alpha plus delta. Microhardness tests were used to identify the various phases. Also, thickness measurements were made of each component in the interface. The normal pressure bonded interface consists mainly of epsilon and delta phases. A cast bonded sample, subjected to the pressure bonding treatment, consists predominantly of alpha plus delta material while an interface bonded during casting is primarily alpha plus delta phases and epsilon phase.

In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).