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The {open_quotes}O{close_quotes} ring seal has been proposed to replace the van stone flange and the bellows thermal expansion assembly currently used on the existing Hanford piles to achieve water and gas seals, respectively. Possible advantages of the {open_quotes}O{close_quotes} ring seal are: (1) simplification of component parts and elimination of van stone corrosion; (2) simplification of maintenance; (3) lower costs of initial erection; (4) increased strength. This test supplements Test Project No. 27 (a preliminary thermal cycling test) in applying the {open_quotes}O{close_quotes} ring seal assembly to actual pile operating conditions.

The Fuel Development operation at Hanford uses a variety of in-reactor facilities to test experimental and prototypical fuel elements. High pressure-high temperature water loops in constant use are the 6 in. {times} 9 in. ETR core facility, the 3 in. {times} 3 in. ETR reflector facility and four front-to-rear loops in the 100-K East reactor at Hanford. Low pressure water cooled test facilities in use are located in the MTR the various Hanford reactors. Stainless steel sheathed the thermocouples 1/16-inch diameter insulated with MgO or Al{sub 2}O{sub 3} when properly fabricated and installed, will reliably measure in-reactor temperature up to 1000 C and at least 4000 MWD/T exposure. This report provides a brief description of some of the temperature monitored in-reactor experiments.

The objective of the production test is detailed in this report is to obtain information about the behavior of Zircaloy-2 jacketed rod-and-tube fuel elements during irradiation at high coolant temperatures specifically, this test should provide information about fuel element dimensional stability, suitability of the rod and tube support system, and behavior of the uranium--Zircaloy-2 bond. Four co-extended Zircaloy-2 jacketed rod-and-tube elements, two with welded tube end closure and two with integrally extruded tube end closures, will be irradiated to 2500 MWD/T at jacket surface temperatures up to 270{degree}C in the KER loops. One Zircaloy-2 jacketed seven-rod cluster UO{sub 2} element will also be irradiated.

Early in the development of an extended surface fuel element for use in the NPR, several 7-rod cluster fuel elements were irradiated to determine the dimensional stability of such geometries at high burnups. These elements were fabricated from small diameter uranium rods clad unbonded in stainless steel tubes and assembled in a rod cluster geometry by various support devices. Zircaloy clad fuel rods were not yet available, the stainless steel clad rods therefore served as a suitable material which would withstand high temperature water over a long period of time and maintain relatively high strength properties. The purpose of the irradiation detailed in this report was to determine the effect of high exposure on the swelling, dimensional stability, microstructure, and physical properties of uranium rods restrained unbonded in stainless steel. At the same time, this test was designed to evaluate the effect of fuel rods operating in a cluster geometry, to monitor the central core temperature of the uranium, to determine the stainless steel-uranium interface heat transfer bond coefficient, and to determine the average specific power of the assembled element. Goal exposure for this irradiation test was 3500 MWD/t.

This report describes travel to Los Alamos Scientific Laboratory September 17, and 19, ALOO and ACF Industries, Inc. At Los Alamos the principal subjects discussed were reservoir design, and reservoir shipping specifications. In Albuquerque a meeting was held with Albuquerque Operations Office personnel to discuss design and procurement of shipping containers, followed by discussions with ACF of reservoir fabrication and testing.

When a ``K`` type process tube is operated without a slug column, damage may take place in the connector because of the cavitation that originates in the venturi. Reported here is a study of this cavitation problem as related to CG-654 operating conditions. A series of curves were developed to present the effect of the venturi diameter on the venturi {Delta}P for instrument operation and on the cavitation problem at the higher flows which are experienced when the tube is operated without a slug column.

It was agreed at the February, 1951 meeting of the Atomic Energy Commission Technical Information Panel that, wherever practical, progress reports should periodically include a list of previous reports issued in the same series. Such a list would greatly facilitate servicing by installation document rooms since requests are frequently received for reports in a specific series. This report lists the Technical Progress Letters from 1944 to date, and is distributed to recipients of the Letters in accordance with the recommendations of the Panel. Because of its length it was not deemed practical to include it as a part of the report itself. lt is anticipated that it will serve as a reference list and permit reduction of personal files.

In the reference letter and later letters the author made recommendations pertaining to the programs of AEC National Laboratories and other AEC contractors in direct support of the Savannah River project during FY-1956. He believes that these recommendations should now be revised in view of schedule changes ad in the light of accumulated experience. The estimated cost of the programs recommended earlier totaled $2,083,000 for FY-1956. The estimated cost of programs currently recommended total $2,048,000 for FY-1956. The programs mentioned in succeeding paragraphs are concerned mainly with the development of processes for producing extended surface fuel elements and, to a lesser extent, with the development of separations processes. The author considers these particular supporting programs important to the rapid and successful achievement of development objectives at the Savannah River Plant and accordingly recommend that the Commission sustain them. Two of the programs, specifically the work by Sylvania and by Horizons to develop processes for reducing thorium oxide to metal, are not now associated with meeting future production objectives at the Savannah River Plant. The author believes that these programs should be brought to an orderly conclusion, as previously planned, to avoid any loss of the art developed to date.

Prior to fiscal year 1955 waste tank scheduling for Separations Plants included the assumptions that T Plant would cease operations in March, 1955 and B Plant would nto be reactivated. At the beginning of the fiscal year these were changed to the assumptions that T Plant would operate through June, 1958 and B Plant, after reactivation, would operate July, 1955 through June, 1957. It has been necessary to alter waste tank scheduling accordingly. The letter of July 23, 1954, {open_quotes}B Plant Waste Storage Forecast{close_quotes}, outlined the plan for the disposal of wastes from B Plant on the assumption that direct feed of metal waste from B Plant ot the TBP Plant would not be done. On July 30, 1954 this letter was reissued as {open_quotes}Waste Storage for B Plant Reactivation{close_quotes}, with a revision including more details and a provision for the direct feed of metal waste. On August 6, 1954 the letter {open_quotes}T Plant First Cycle Waste{close_quotes} was issued recommending that scavenging of T Plant first cycle waste be started by December 1, 1954. On October 22, 1954 the letter entitled {open_quotes}Justification for T Plant First Cycle Waste Scavenging{close_quotes} gave the detailed economic justification. The present letter summarizes the preceding …

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The composition of dissolver off-gas was determined for seven different dissolvings of uranium in nitric acid (both with one and two dissolvers in operation) at irregular intervals during February, March, and April, 1958. Samples were taken at the Purex 293-A facility absorber inlet, scrubber inlet, and scrubber discharge. The absorber was operated with and without reflux, and the scrubber was operated with water or with approximately 10% sodium hydroxide. This memorandum describes the sampling procedures and analytical methods and gives the gas analyses found.

A review is given of the various transverse thermomagnetic and galvanomagnetic effects. The coefficicnts and relations between coefficients are listed for both metals and semiconductors. A quantitative annlysis is given of the error in Hall effcct measurements owing to non-isothermal conditions. (auth)

This report details activities of the Nuclear Physics Research Operation for the month of August 1958.

In this report, the equation (HW-55219-RD) expressing tube-wise rupture rate as a function of exposure, power, and temperature has been adjusted to include 1958 data.

This production test was designed to evaluate the suitability of low hydrogen dingot uranium as routine process material. Nine tubes of I and E fuel elements (6 dingot, 3 ingot) with 32 fuel elements in each tube, have recently been discharged at the C Reactor and this document contains the results of analyses made on the dimensional stability properties of this material.

Attached is a draft of revisions being made to subject document, HW-53612. These revisions pertain to the moderators only, and will be incorporated in an overall revision of the technical criteria, which will be issued soon. Requirements, criterion, and data are included.

This report discusses the events from a chemical standpoint following a total loss of coolant disaster will not be altered in the melting reactor by the introduction of N metal. The interdiffusion of uranium and aluminum will be the dominating reaction, causing the blockage and tying up of the lithium in UAl{sub 3} which does not melt until after the uranium does. Pressure from the swelling UAl{sub 3} will extrude uranium-aluminum and lithium into graphite weep holes and block interfaces. The migration of lithium by vaporization will not became appreciable until well over 2000{degrees}C, well beyond the time when uranium and UAl{sub 3} have melted. The eventual result will be a diffuse distribution of uranium, lithium, and aluminium in the lattice. The E-N pile has a larger excess over required control capacity than the uranium provided the large reactivity poison tied up in the lithium is not lost. Compared to the natural uranium pile, the gain of reactivity on loss of coolant is less and the net temperature coefficient in the dry pile remains negative to higher exposures. Furthermore, permanent pile poisoning during meltdown is accomplished via two mechanisms both lithium and uranium redistribution in the lattice produce large negative …

The purpose of this report is to summarize briefly the work which has been done to date and to acquaint the reader with the results of the ``Hot Spot Investigation Program `` as well as some of the problems encountered.

Neptunium-237 is currently produced in the Hanford reactors at a rate of approximately .003 gms/MWD via the following reactions: (a) U{sup 238} (n,2n) U{sup 237}{sup {beta}}{yields} Np{sup 237}. (b) U{sup 235} (n,{gamma}) U{sup 236} (n,{gamma}) U{sup 237}{sup {beta}}{yields} Np{sup 237}. In order to calculate the buildup of Np{sup 237} via reaction (a), which accounts for the greater share of the formation of Np{sup 237}, the n,2n cross section for U{sup 238} must be known. An old value quoted by Arnold of 5.2 millifermis for an {open_quotes}effective{close_quotes} 2200 m/s value is not large enough to account for the observed Np{sup 237} yield by about a factor of two. Recent n,2n cross section measurements for U{sup 238} permit a newer calculation and the result is 11.2 mF, effective 2200 m/s value.

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